ORAU Team 

NIOSH Dose Reconstruction Project

Technical Basis Document for the Y-12 National Security Complex

– Occupational External Dosimetry 

Document Number: 

ORAUT-TKBS-0014-6 

Effective Date: 11/19/2003 

Revision No.: 00 

Controlled Copy No.:_______ 

Page 1 of 47

Subject Expert: George D. Kerr 

Document Owner 

Approval: Signature on File Date: 11/19/2003 William E. Murray, TBD Team Leader 

Approval: Signature on File Date: 11/19/2003 Judson L. Kenoyer, Task 3 Manager 

Concurrence: Signature on File Date: 11/19/2003 Richard E. Toohey, Project Director 

Approval: Signature on File Date: 11/21/2003 James W. Neton, OCAS Health Science Administrator 

Supersedes: 

None

TABLE OF CONTENTS 

Section Page

Record of Issue/Revisions......................................................................................................................5

Acronyms and Abbreviations..................................................................................................................6

6.1 Introduction ....................................................................................................................................8

6.2 Basis of Comparison......................................................................................................................9

6.3 Dose Reconstruction Parameters................................................................................................10

6.3.1 Y-12 Historical Administrative Practices .............................................................................10

6.3.2 Y-12 Dosimetry Technology................................................................................................13

6.3.2.1 Beta/Photon Dosimeters ............................................................................................13

6.3.2.2 Neutron Dosimeters ...................................................................................................16

6.3.3 Y-12 Dosimeter Calibration Procedures .............................................................................17

6.3.3.1 Beta/Photon Dosimeters ............................................................................................17

6.3.3.2 Neutron Dosimeters ...................................................................................................18

6.3.4 Y-12 Workplace Radiation Fields........................................................................................19

6.3.4.1 Workplace Beta/Photon Dosimeter Response...........................................................19

6.3.4.2 Workplace Neutron Dosimeter Response..................................................................21

6.3.4.2.1 Calibration Laboratory in Building 9983..............................................................21

6.3.4.2.1.1 Neutron Energy Spectrum ........................................................................22

6.3.4.2.1.2 Neutron-to-Photon Dose Ratio .................................................................23

6.3.4.2.2 Enriched Uranium Storage Area in Building 9212 ..............................................23

6.3.4.2.2.1 Neutron Energy Spectrum ........................................................................23

6.3.4.2.2.2 Neutron-to-Photon Dose Ratio .................................................................24

6.3.4.2.3 Nondestructive Assay Laboratory in Building 9720-5 .........................................25

6.3.4.2.3.1 Neutron Energy Spectrum ........................................................................25

6.3.4.2.3.2 Neutron-to-Photon Dose Ratio .................................................................26
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 2 of 47 

6.3.4.3 Typical Workplace Neutron Dosimeter Hp(10) Performance .....................................26

Section Page

6.4 Adjustments of Recorded Dose ...................................................................................................27

6.4.1 Photon Dose Adjustments ..................................................................................................27

6.4.2 Neutron Dose Adjustments .................................................................................................28

6.4.2.1 Neutron Weighting Factors ........................................................................................29

6.4.2.2 Neutron Dose Correction Factors ..............................................................................29

6.5 Missed Dose ................................................................................................................................30

6.5.1 Photon Missed Dose...........................................................................................................30

6.5.1.1 Facility or Location .....................................................................................................31

6.5.1.2 Dosimeter Type..........................................................................................................31

6.5.1.3 Year............................................................................................................................31

6.5.1.4 Energy Range ............................................................................................................31

6.5.2 Neutron Missed Dose .........................................................................................................32

6.5.2.1 Prior to 1980...............................................................................................................32

6.5.2.2 From 1980 to 1989.....................................................................................................32

6.5.2.3 Post 1989 ...................................................................................................................33 

6.6 Uncertainty in Photon and Neutron Dose ....................................................................................33

6.7 Organ Dose .................................................................................................................................33

References ...........................................................................................................................................35

Glossary................................................................................................................................................40

Attachment F Occupational External Dose For Monitored Workers 

F.1 Occupational External Dose ........................................................................................................41

F.2 Years of Exposure .......................................................................................................................41

F.3 Exposure Rate .............................................................................................................................41

F.4 Radiation Type.............................................................................................................................41

F.4.1. Beta and Photon Radiation .................................................................................................41

F.4.2. Neutron Radiation ...............................................................................................................42

F.5 Adjustments to Recorded Dose ...................................................................................................43

F.5.1. Parameter #1 ......................................................................................................................43

F.5.2. Adjustment to Recorded Photon Dose................................................................................43

F.5.3. Adjustments to Recorded Neutron Dose ............................................................................43

F.5.4. Unmonitored Photon and Neutron Dose.............................................................................44

F.5.5. Missing Photon Dose ..........................................................................................................44

F.5.5.1. Facility or Location .....................................................................................................44

F.5.5.2. Dosimeter Type..........................................................................................................45

F.5.6. Neutron Missed Dose .........................................................................................................45

F.5.7 Organ Dose Equivalent.......................................................................................................46

F.5.8 Parameter #2 ......................................................................................................................46
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 3 of 47 

F.5.9 Organ Dose Conversion Factors.........................................................................................46

LIST OF TABLES 

6.3.1-1 Monitoring technique and exchange frequency used at Y-12 Plant for 

external whole-body exposures ..........................................................................................11

6.3.1-2 Dosimeter type, period of use, exchange frequency, laboratory 

minimum detectable limit, and maximum annual missed dose...........................................12

6.3.2-1 Y-12 historical dosimetry events .........................................................................................14

6.3.2.1-1 IARC study results for US beta/photon dosimeters.............................................................16

6.3.3.1-1 Common sources of laboratory bias in the calibration parameters for 

beta/photon dosimeters ......................................................................................................18

6.3.3.2-1 Common sources of laboratory bias in the calibration parameters for 

neutron dosimeters .............................................................................................................19

6.3.4.1-1 Selection of beta and photon radiation energies and percentages for 

Y-12 site processes ............................................................................................................20

6.3.4.1-2 Typical workplace beta/gamma dosimeter Hp(10) performance.........................................21

6.3.4.2-1 Dose fractions for Y-12 calibration laboratory.....................................................................23

6.3.4.2-2 Neutron dose fractions for Y-12 enriched uranium storage area ........................................24

6.3.4.2-3 Dose fractions for Y-12 nondestructive analysis laboratory................................................26

6.3.4.3-1 Typical workplace neutron dosimeter Hp(10) performance ................................................26

6.3.4.3-2 Number of neutron monitored workers, cumulative neutron dose, and 

average neutron dose to Y-12 workers for 12-year period following 

introduction of the four-element TLND dosimeter in 1989 ..................................................27

6.4.1-1 Adjustments to reported Y-12 deep photon dose ...............................................................28

6.4.2.1-1 Neutron quality factor, Q, or weighting factor, wr ................................................................30

6.4.2.2-1 Summary of neutron dose fractions and associated ICRP 60 

correction factors for Y-12 site facilities ..............................................................................30

6.6-1 Default exposure geometries for calculating organ doses ..................................................34

F-1 IREP dose parameter input screen.....................................................................................41

F-2 Selection of beta and photon radiation energies and percentages for 

Y-12 site processes ............................................................................................................42

F-3 Selection of neutron energies and percentages for Y-12 site facilities ...............................42

F-4 Adjustments to reported Y-12 deep photon dose ...............................................................43

F-5 Neutron dose fractions and associated ICRP 60 correction factors for 

Y-12 site facilities ................................................................................................................45 

F-6 Dosimeter type, period of use, exchange frequency, laboratory 

minimum detectable limit, and maximum annual missed dose...........................................45

F-7 Default exposure geometries for calculating organ dose....................................................46
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 4 of 47 

LIST OF FIGURES 

6.1-1 Percentage of Y-12 workers monitored ................................................................................9

6.3.2.1-1 Comparison of Hp(10) for photons......................................................................................15

6.3.2.2-1 Comparision of Hp(10) for neutrons....................................................................................17

6.3.4.2-1 Results of PNL neutron spectrum measurements in Calibration 

Laboratory of Building 9983 ................................................................................................22

6.3.4.2-2 Results of PNL neutron spectrum measurements in Enriched Uranium 

Storage Area.......................................................................................................................24

6.3.4.2-3 Results of PNL neutron spectrum measurements made at 1 meter 

from a bare 252Cf fission neutron source .............................................................................25

6.4.1-1 Maximum and average photon deep dose to Y-12 workers for the 10-

year period from 1978 to 1987............................................................................................28

6.4.2.1-1 Comparison of the neutron quality factors used in the PNL neutron 

spectrum measurements and the neutron quality factors from NCRP 

Report 38 ............................................................................................................................29
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 5 of 47 

RECORD OF ISSUE/REVISIONS 

ISSUE 

AUTHORIZATION 

DATE 

EFFECTIVE 

DATE 

REV. 

NO. DESCRIPTION 

Draft 10/13/2003 00-A New document to establish occupational external 

dosimetry for the Y-12 National Security Complex. 

Initiated by William E. Murray. 

11/19/2003 11/19/2003 00 First approved issue. Initiated by William E. 

Murray. 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 6 of 47 

ACRONYMS AND ABBREVIATIONS 

AEC U.S. Atomic Energy Commission 

ANSI American National Standards Institute 

A-P Anterior-posterior (or front-to-back) irradiation of the body 

BXWT Y-12 Current operator of Y-12 under contract with DOE 

CATI Computer assisted telephone interview (of claimant) 

Ci Curie (a unit of radioactivity) 

DCF Dose conversion factor 

DOE U.S. Department of Energy 

DOELAP DOE Laboratory Accreditation Program 

DU Depleted uranium 

EEOICPA Energy Employees Occupational Illness Compensation Program Act 

ETTP East Tennessee Technological Park 

EU Enriched uranium 

HPRR Health Physics Research Reactor 

IAEA International Atomic Energy Agency 

IARC International Agency for Research on Cancer 

ICRP International Commission on Radiological Protection 

ICRU International Commission on Radiation Units and Measurements 

IREP Interactive Radio-Epidemiological Program (a computer program) 

ISO International Standards Organization 

K-25 Oak Ridge Gaseous Diffusion Plant, now the ETTP 

keV kiloelectronvolt (a unit of energy) 

MDL Minimum Detectable Limit 

MED Manhattan Engineering District 

MeV Megaelectronvolt (a unit of energy) 

MMES Martin Marietta Energy Systems 

NCRP National Council on Radiation Protection and Measurements 

NIOSH National Institute for Occupational Safety and Health 

NTA Nuclear track type A emulsion (or film) 

ORGDP Oak Ridge Gaseous Diffusion Plant 

ORISE Oak Ridge Institute for Science and Education 

ORNL Oak Ridge National Laboratory 

PC Probability of causation 

PIC Pocket ionization chamber 

PNL Pacific Northwest Laboratory 

RADCAL Radiation Calibration Laboratory at ORNL 

REF Radiation Effectiveness Factor 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 7 of 47 

rem Roentgen equivalent man (a unit of radiation dose equivalent) 

RPG Radiation protection guideline 

SD Standard deviation 

SSN Social Security Number 

TEC Tennessee Eastman Corporation 

TLD Thermoluminescent Dosimeter 

TLND Neutron TLD 

WB Whole body 

UCC Union Carbide Corporation 

UCC-ND UCC Nuclear Division 

Y-12 Y-12 Plant, now the Y-12 National Security Complex 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 8 of 47 

6.1 INTRODUCTION 

The Y-12 Plant, now the Y-12 National Security Complex, was first conceived in the fall of 1942 by 

engineers of the Manhattan Engineering District (MED) of the U.S. Army Corps of Engineers, and the 

construction of the first building was completed in 1943 (Wilcox 2001). The Tennessee Eastman 

Corporation (TEC) operated Y-12 between June 1943 and May 1947. During this period, the 

operations at Y-12 primarily involved the use of the electromagnetic separation process to enrich 

uranium in 235U, with the enriched product being shipped to Los Alamos for production of atomic 

weapons. Until the latter part of 1945, Y-12 converted UO3 to UCl4 which was subsequently enriched 

in 235U by the electromagnetic separation process using two calutron stages (termed “alpha” and 

“beta”). In late 1945, Y-12 discontinued the use of the alpha calutron stage, and processes at Y-12 

were changed to receive UF6 from the Oak Ridge Gaseous Diffusion Plant (ORGDP) or so-called K-25 

Plant. The UF6 was then further enriched at Y-12 by the beta calutrons and shipped to Los Alamos. 

In these early days of Y-12, TEC relied entirely on facility monitoring to measure and control the 

radiation exposure to workers. The nature of the work at Y-12 in these early years resulted in internal 

occupational exposure being more important than occupational external exposure. 

In May 1947, management of Y-12 was assigned to the Carbide & Chemicals Company, a division of 

the Union Carbide & Carbon Corporation, and emphasis was directed away from enrichment to the 

fabrication of nuclear weapon parts. Numerous changes have occurred over the years in the 

fabrication procedures, but the general features have remained the same. Typically, enriched 

uranium (EU) was received at Y-12 in the form of UF6, converted to UF4, reduced to a metal, and then 

fabricated into weapon parts. These fabrication processes involved casting of metal, rolling and 

forming the metal, machining the metal, and recycling of the EU salvage. The fabrication of weapon 

parts was expanded over the years to include other radioactive and non-radioactive materials. In 

addition to facility monitoring to measure and control the radiation exposure to workers, an external 

dosimetry program was started in 1948 to monitor individual personnel working in the Assay 

Laboratories, Radiographic Shop, Spectrographic Shop, and the “Metal” Machine Shops. This 

program which monitored less than 25% of the total number of Y-12 employees was continued 

through the criticality accident at Y-12 in 1958. As a result of the 1958 criticality accident, a program 

was instituted in 1961 to individually monitor all Y-12 workers for external radiation exposure using a 

dosimeter system that was an integral part of the worker’s identification badge and contained 

components for both routine and accident dosimetry. Thus, Y-12 has used both facility monitoring 

and individual worker monitoring to measure and control radiation exposures to radiation workers 

since 1948. The percentages of Y-12 workers monitored for external radiation exposure from the start 

of the external dosimetry program in 1948 through the switch to monitoring nearly all workers in 1961 

are shown in Figure 6.1-1. The external monitoring data for Y-12 workers from 1948 to 1950 are not 

readily available by Social Security Number (SSN) and are not being supplied by Y-12 in response to 

Energy Employees Occupational Illness Compensation Program Act (EEOICPA) requests 

(Souleyrette 2003). 

There are numerous Y-12 records concerning facility monitoring, safety evaluations, investigations, 

etc. However, it is time consuming to locate and evaluate these records for all Y-12 facilities and 

processes since 1943. Evaluations of the extensive scope of facility, process, and worker information 

relevant to an individual worker’s potential dose, many years or even decades after employment, are 

difficult or even impossible in some instances. Records of radiation dose to individual workers from 

personnel dosimeters worn by the worker and co-workers are available for the employees with the 

highest potential for external radiation exposure from 1950 to 1961 and for all workers from 1961 to 

the present. Dose from these dosimeters is recorded at the time of measurement, reviewed routinely 

by Y-12 operations and safety staff for compliance with radiation control limits, and made available 

routinely to workers. The National Institute for Occupational Safety and Health (NIOSH) External 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 9 of 47 

Dosimetry Implementation Guide (NIOSH 2002) has identified these records to represent the highest 

quality record to retrospectively assess dose. Information presented in this section pertains to 

1945 1950 1955 1960

Year

0

5

10

15

20

25

External Monitoring of Y-12 Workers (%)

Figure 6.1-1. Percentage of Y-12 workers monitored for external 

radiation exposure from start of external monitoring program in 

1948 through the switch to monitoring nearly all workers in 1961 

(Watkins et al. 1993). 

the analysis of these records and does not address parameters regarding skin and testicular or breast 

radiation dose that may result from acute exposure to beta-particles in generally non-routine 

workplace exposure situations. 

Radiation dosimetry practices were initially based on experience gained during several decades of 

radium and x-ray usage in medical diagnostic and therapeutic applications. These methods were 

generally well advanced at the start of the MED project to develop nuclear weapons in the 1940s 

(Morgan 1961; Taylor 1971). The primary difficulties encountered in MED efforts to measure worker 

doses to external radiation were (1) the large quantities of high level radioactivity that were not 

encountered previously and (2) the mixed radiation fields involving beta particles, photons (x-rays and 

gamma rays), and neutrons having a broad spectrum of energies. 

6.2 BASIS OF COMPARISON 

Since the initiation of the MED in the early 1940s, various radiation dose concepts and quantities 

have been used to measure and record occupational dose. A basis of comparison for dose 

reconstruction is the Personal Dose Equivalent, Hp(d), where d identifies the depth (in mm) and 

represents the point of reference for dose in tissue. For weakly penetrating radiation of significance to 

skin dose, d = 0.07 mm and is noted as Hp(0.07). For penetrating radiation of significance to “whole 

body” dose, d = 10 mm and is noted as Hp(10). Both Hp(0.07) and Hp(10) are the radiation quantities 

recommended for use as the operational quantity to be recorded for radiological protection purposes 

by the International Commission on Radiation Units and Measurements (ICRU 1993). In addition, 

Hp(0.07) and Hp(10) are the radiation quantities used in the U.S. Department of Energy’s (DOE) DOE 

Laboratory Accreditation Program (DOELAP) to accredit DOE personnel dosimetry systems since the 

1980s (DOE 1986). The International Agency for Research on Cancer (IARC) also selected Hp(10) 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 10 of 47 

as the quantity to assess error in historical, recorded “whole body” dose for workers in the IARC 

nuclear worker epidemiologic studies (Thierry-Chef et al. 2002). The basis for comparison for neutron 

radiation is more complicated since historically the calibration of dosimeters to measure neutron dose 

was based on different dose quantities such as First Collision Dose, Multiple Collision Dose, Dose 

Equivalent Index, etc. The numerical difference in using these dose quantities compared to the 

Hp(10) dose used in current DOELAP performance testing could be evaluated by using the relative 

values of the dose conversion factors for the respective dose quantities in conjunction with 

characteristics of the respective Y-12 neutron dosimeter response characteristics and workplace 

radiation fields. 

6.3 DOSE RECONSTRUCTION PARAMETERS

Examinations of the beta, photon (x-ray and gamma ray), and neutron radiation types, energies and 

exposure geometries, and the characteristics of the respective Y-12 dosimeter response are crucial 

for the assessment of bias and uncertainty of the original recorded dose in relation to the radiation 

quantity Hp(10). The bias and uncertainty for current dosimetry systems are typically well 

documented for Hp(10). Often the performance of current dosimeters can be compared with 

performance characteristics of historical dosimetry systems in the same, or highly similar, facilities or 

workplaces. In addition, current performance testing techniques can be applied to earlier dosimetry 

systems to achieve a consistent evaluation of all dosimetry systems. Dosimeter response 

characteristics for radiation types and energies in the workplace are crucial to the overall analysis of 

error in recorded dose. 

Overall, the accuracy and precision of the original recorded individual worker doses and their 

comparability to be considered in using NIOSH (2002) guidelines depend on: 

• Administrative practices adopted by facilities to calculate and record personnel dose based 

on technical, administrative, and statutory compliance considerations. 

• Dosimetry technology used that includes the physical capabilities of the dosimetry system, 

such as the response to radiation type and energy, especially in mixed radiation fields. 

• Calibration methods used for the respective monitoring systems and the similarity of the 

methods of calibration to sources of exposure in the workplace. 

• Workplace radiation fields that may include mixed types of radiation, variations in exposure 

geometries, environmental conditions. 

An evaluation of the original recorded doses based on these parameters is expected to provide the 

best estimate of Hp(0.07), as necessary, and Hp(10) for individual workers, with the least relative 

overall uncertainty. 

6.3.1 Y-12 Historical Administrative Practices

A dosimetry program was started in 1948 to monitor individual external exposures of personnel 

working in the Assay Laboratories, Radiographic Shop, Spectrographic Shop, and “Metal” Machine 

Shops. At first, the external radiation monitoring was performed using pocket ionization chambers 

(PICs), typically exchanged on a weekly basis (Souleyrette 2003). Additionally, early efforts were 

concerned with using a photographic film pad on the hands of the uranium metal workers and 

attempting to correlate the film pad reading with whole-body exposures, which were recorded first with 

PICs and later with personnel whole-body (WB) film badge dosimeters. The Y-12 film badge program 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 11 of 47 

for external monitoring of exposures to beta particles, photons (x-rays and gamma rays), and 

neutrons continued through 1980, when the personnel film dosimeters were replaced with 

thermoluminescent dosimeters (TLDs). The frequency of exchange of these personnel dosimeters is 

summarized in Table 6.3.1-1 (Souleyrette 2003; Watkins et al. 1993; West 1993). 

Table 6.3.1-1. Monitoring technique and exchange frequency used at Y-12 Plant for external wholebody exposures.a

Period Monitoring technique Exchange frequency Monitored personnel

Beta/photon dosimeters 

1948-1950 Pocket ionization 

chambers and twoelement film dosimeters 

Some daily, some weekly Personnel expected to receive 

over 10% of RPG 

1950-04/07/58 Two-element film 

dosimeters 

Weekly Personnel expected to receive 

over 10% of RPG 

04/08/58-06/30/61 Two-element film 

dosimeters 

Monthly Personnel expected to receive 

over 10% of RPG 

06/30/61-10/01/80 Four-element film 

dosimeters 

Quarterly Nearly all personnel 

monitored 

10/01/80-01/03/89 Two-element TLD 

dosimeters 

Some quarterly, some 

annually, a very limited 

group on a monthly basis 

Quarterly exchange for 

personnel expected to receive 

500 mrem or more, annual 

exchange for personnel 

expected to receive less than 

500 mrem 

01/03/89-Present Four-element TLD 

dosimeters 

Mostly quarterly, some 

monthly 

Nearly all personnel 

monitored from 1989 to 1996. 

After 1996, only personnel 

entering radiological areas. 

Neutron dosimeters 

1950-10/01/80 NTA film Biweekly, monthly, and 

quarterly 

Personnel exposed to neutron 

sources 

10/01/80-01/03/89 NTA film for fast neutrons 

and TLND dosimeters for 

other energy neutrons 

Quarterly Personnel exposed to 

neutron sources 

01/03/89-Present TLND dosimeters for 

neutrons of all energies 

Quarterly Personnel exposed to 

neutron sources 

a. Souleyrette (2003), Watkins et al. (1993), West (1993). 

The minimum detection level (MDL) of the various dosimeters used at Y-12 to monitor for 

beta/gamma and neutron exposures of the whole body is summarized in Table 6.3.1-2. The first film 

dosimeter used at Y-12 is believed to be the same badge used at the Oak Ridge National Laboratory 

(ORNL) in 1949 (West 1993) and described by Thornton, Davis, and Gupton (1961). This film badge 

was an AEC Catalog Number PF-1B film badge manufactured by the A. M. Samples Machine 

Company in Knoxville, Tennessee (Patterson, West and McLendon 1957). It had an open window 

over a portion of the film to measure both beta and photon radiation, and a 1-mm thick cadmium filter 

over a portion of the film to measure higher energy photon radiation only. This film dosimeter was 

used at Y-12 until 1961, when a newer film dosimeter was developed for use at all Union Carbide 

Corporation Nuclear Division (UCC-ND) facilities (Thornton, Davis, and Gupton 1961). This newer 

film dosimeter served as identification badge and also provided for both personnel routine and 

accident monitoring. The precise minimum detectable limits (MDLs) shown in Table 6.3.1-2 are 

difficult to estimate, particularly for the film dosimeters. For the current TLD dosimeters, the MDLs are 

precisely identified in the Y-12 External Dosimetry Technical Basis Documentation (BWXT Y-12 2001) 

based on a DOELAP protocol (DOE 1986). For the film dosimeters, the MDLs are subject to 

additional uncertainty because factors involving the radiation field and film type, as well as the 

processing, developing, and reading system cannot be tested. The MDLs for the film dosimeters in 

Table 6.3.1-2 were based on information from 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 12 of 47 

Table 6.3.1-2. Dosimeter type, period of use, exchange frequency, laboratory minimum detectable 

limit, and maximum annual missed dose.a

Dosimeter Period

Exchange 

frequency

Laboratory 

MDL (mrem)

Maximum annual 

missed dose 

(mrem)

Beta/photon dosimeters 

Pocket ionization chamber 1948-1950 Daily < 5 1,300 

Weekly < 5 260 

Two-element film badge 1948-1958 Weekly 40 2,080 

1958-1961 Monthly 40 480 

Four-element film badge 1961-1980 Quarterly 40 160 

Two-element TLD dosimeter 1980-1989 Quarterly 20 80 

Four-element TLD dosimeter 1989-Present Quarterly 10 40 

Neutron dosimeters 

Biweekly < 50 1,300b

Monthly < 50 600b

NTA film 1948-1980 

Quarterly < 50 200b

1980-1985 Quarterly < 50 200b Combination NTA film and TLND 

dosimeter 1985-1989 Quarterly 20 80 

TLND dosimeter 1989-Present Quarterly 10 40 

a. Souleyrette (2003), Watkins et al. (1993), West (1993), Wilson et al (1990), Patterson, West, and McLendon (1957). 

b. Potential annual missed dose based on data from laboratory irradiations may not be directly applicable to workplace missed dose. 

Souleyrette (2003), Watkins et al. (1993), West (1993), Wilson et al. (1990), and Patterson, West, and 

McLendon (1957). Y-12 administration practices that are important to dose reconstruction include the 

following policies for: 

• Assigning dosimeters to workers 

• Exchanging dosimeters 

• Recording notational dose (i.e., some identified values for lower dose workers based on a 

small fraction of the regulatory limit) 

• Estimating dose from lost or damaged dosimeters 

• Replacing destroyed or missing records 

• Evaluating and recording doses for incidents or accidents 

• Obtaining and recording occupational dose to workers for other employer exposures. 

Policies appear to have been in place at Y-12 for all of these parameters. Their routine practices 

appear to have required assigning dosimeters to personnel who might receive an external radiation 

dose that was greater than 10% of the Radiation Protection Guidelines (RPGs) that was in effect at 

that time. Dosimeters were exchanged on a routine schedule. All beta/photon dosimeters were 

processed and the measured results recorded and used for dose estimation to the individual workers. 

A neutron thermoluminescent dosimeter (TLND) dosimeter was not issued to a worker or the nuclear 

track type A (NTA) emulsion in a film badge dosimeter was not read unless a worker was exposed to 

neutrons. There appears to be no use of recorded notional doses, although there are issues of 

recording dose for low-dose exposures (see Table 6.3.1-2). There is also the problem of missing 

dose components from the total WB dose for a worker designated simply as “not available” or 

“damaged film” in the worker’s records (West 1993). These missing dose components for workers 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 13 of 47 

could be estimated using a method described by Watson et al. (1994) and based on examination of 

continuity in the worker’s job and work activities. 

6.3.2 Y-12 Dosimetry Technology

The Y-12 dosimetry methods evolved during the years as improved technology was developed and 

the complex radiation fields encountered in the workplace were better understood. The adequacy of 

the respective dosimetry methods to accurately measure radiation dose as discussed in later sections 

depends on radiation type, energy, exposure geometry, etc. The dosimeter exchange frequency of 

the dosimeters was gradually lengthened and corresponded generally to downward reductions in the 

RPGs (Morgan 1961; Taylor 1971). During the early stages of the Y-12 program to monitor individual 

workers, a weekly dose control of 0.3 rem was in effect. This was changed to an annual limit of 5 rem 

in the latter part of the 1950s. A summary of the major historical events in the Y-12 dosimetry 

program for external radiation is provided in Table 6.3.2-1. 

6.3.2.1 Beta/Photon Dosimeters 

In 1948, Y-12 started an external dosimetry badge service with the assistance of ORNL (Murray 

1948a, b). The ORNL had earlier implemented the beta/photon film dosimeter design that was 

developed originally at the Metallurgical Laboratory of the University of Chicago (Pardue, Goldstein, 

and Wollan 1944). Several minor modifications had been made to this original design as discussed 

by Patterson, West, and McLendon (1957) and by Thornton, Davis, and Gupton (1961). The film 

badge was a so-called two element badge because a portion of the film was covered with a 

comparatively “open window” and a portion of the film was covered by a cadmium shield or filter. In 

1961, the two-element film badge was replaced by a multi-element film badge with an “open window” 

over a portion of the film and three filters of plastic, aluminum, and cadmium over other portions of the 

film (Thornton, Davis, and Gupton 1961; McLendon 1963; McRee, West, and McLendon 1965). 

Cadmium filters have been consistently incorporated in all film badge designs used at Y-12 since 

1948 and have an approximate thickness of 1 mm or mass density of 1000 mg cm-2. The external 

doses to Y-12 workers from photons were always determined from film readings behind the cadmium 

filter (Sherrill and Tucker 1973). In addition, the Y-12 film badges have always included a 

comparatively “open window” to measure significant beta radiation and to distinguish film exposures 

due to beta and photon radiation. The film areas behind the plastic and aluminum filters of the multielement film dosimeters were read at Y-12, but they were not used in the normal evaluation of worker 

doses (Sherrill and Tucker 1973). 

The film badges used at Y-12 from 1948 to 1963 contained DuPont type 552 film packets (Souleyrette 

2003). These packets consisted of two film emulsions: (1) a sensitive 502 emulsion with an effective 

range from approximately 30 mrem to 10 rem, and (2) an insensitive 510 emulsion with an effective 

range from approximately 500 mrem to 20 rem (Craft, Ledbetter, Hart 1952; Thornton, Davis and 

Gupton 1961; Parrish 1979). In 1963, Y-12 switched to the use of DuPont 554 film packets 

(McLendon 1963; Souleyrette 2003). These packets consisted of the following two film emulsions: (1) 

a sensitive 555 emulsion with an effective range from approximately 30 mrem to 5 rem (McLendon 

1963), and (2) an insensitive 834 emulsion with an effective range from approximately 5 rem to 150 

rem (Thornton, Davis, and Gupton 1961; Davis and Gupton 1963; Parrish 1979). In 1971, DuPont 

stopped manufacturing the 554 film packets and Y-12 switched to Eastman type 2 film (Jones 1971; 

Souleyrette 2003). Eastman type 2 packet contained one film with two emulsions bonded to opposite 

side of the base film. The sensitive emulsion had a minimum detectable limit of approximately 30 

mrem (Jones 1971). During the switch to the Eastman type 2 film, some film to be evaluated was 

removed from cold storage, inserted in 16 pairs of badges, and the badges placed in racks with the 

office area to investigate on-site radiation background. A sample of film was developed on the day of 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 14 of 47 

Table 6.3.2-1. Y-12 historical dosimetry events. 

Date Event Reference

1948 An external dosimetry program was started to monitor personnel in 

radiological areas using pocket ionization chambers (PICs) and twoelement film badges. Film pads were also used as an extremity monitor 

for beta-particle exposure to the hands of uranium “metal” workers. 

Murray 1948a,b; 

Struxness 1949a 

1949 NTA film used to monitor personnel exposed to neutron sources. Struxness 1949b 

1951 Converted from film pads to film rings as an extremity monitor for betaparticle exposure to the hands of uranium “metal” workers. Struxness 1951 

1955 Strips of indium foil with a mass of approximately 1 g each were included 

in the security badges of all employees at Y-12. The foils provided a 

quick means of identifying employees who may have been exposed to 

elevated radiation levels during a nuclear criticality accident. 

McLendon 1959 

June 16, 

1958 

A nuclear criticality accident occurred during recovery of enriched 

uranium in Building 9212 of the Y-12 Plant. The indium foils in the 

security badges of 31 employees indicated that they were exposed to 

elevated levels of thermal neutron radiation during the accident. 

Callihan and Thomas 

1959; McLendon 1959; 

Hurst, Ritchie and 

Emerson 1959 

July 1, 

1961 

All UCC-ND employees were issued a badge meter which served as 

both a security pass and a routine and accident dosimeter. The badge 

meter contained a four-element film dosimeter for routine monitoring of 

beta particle, photon, and neutron exposures. The badge meter also 

contained indium, other neutron activation foils, and special photon 

dosimeters for nuclear accident dosimetry. 

Thornton, Davis 

and Gupton 1961; 

McLendon 1963; 

McRee, West and 

McLendon 1965 

October

1, 1980 

All UCC-ND employees were issued a badge meter consisting of (1) a 

security badge for identification and (2) a two-element TLD badge for 

personnel radiation monitoring of beta particles and photons. A 1-g foil 

of indium was included in the security badge to provide a quick means of 

identifying personnel exposed to elevated levels of radiation during a 

nuclear accident. An additional ORNL neutron badge containing NTA 

film and TLND dosimeters was issued to Y-12 personnel who were 

exposed to neutrons. 

McLendon 1980a; 

Howell and Batte 1982; 

Gupton 1978; Berger 

and Lane 1985 

1980 The accident dosimetry components from the old “1961" UCC ND badge 

meter were consolidated and incorporated into Y-12 security badge 

covers. All persons were required to use these security badge covers 

when entering controlled areas that had an installed criticality alarm. 

McLendon 1980b; 

Y-12 Plant 1982a 

January

3, 1989 

All MMES employees were issued a badge meter consisting of (1) a 

security badge for identification and (2) a four-element TLD badge for 

personnel radiation monitoring of beta particles and photons. A 1-g foil 

of indium was included in the security badge for accident dosimetry 

purposes, and an additional four-element TLND badge was issued to 

Y-12 personnel potentially exposed to neutrons. 

Y-12 Plant 1988b; 

BWXT Y-12 2001 

1991 The beta-gamma dosimeter (TLD) issued to all MMES employees 

became the official personnel nuclear accident dosimetry (PNAD) in the 

event of a criticality accident. All persons were required to use a betagamma dosimeter attached to their security badge before entering 

controlled areas with an installed criticality alarm. 

Y-12 Plant 1991; Kerr 

and Mei 1993 

the background study and used as a base point for other measurements. Every two weeks, film from 

a pair of badges was unloaded, developed, and read with a densitometer. On the first day of the 

study, the fresh film had an optical density of 0.205, and after 215 days, the optical density had 

reached a level of 0.405. This increase in the optical density of 0.2 represented an increase of only 

about 15 mrem or approximately 0.5 mrem per week. It was noted that these optical density readings 

did not have the optical density readings of an unexposed blank subtracted from them and, therefore, 

represent the actual optical density readings of the films (Jones 1971). 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 15 of 47 

The response of the film badge to photon radiation of different energies is illustrated in Figure 6.3.2.1-

1. This figure also shows the Hp(10) response. Two responses are shown for the film badges: one 

response for a sensitive DuPont 502 emulsion in a two-element film badge (Pardue, Goldstein, and 

Wollan 1944), and one response for a sensitive DuPont 555 emulsion in the a multi-element film 

badge (Thornton, Davis, and Gupton 1961). The response of the sensitive Eastman type 2 film in a 

multi-element film badge should be quite similar to that of the sensitive DuPont 555 emulsion. The 

film badges show an under-response at the lower photon energies and an over-response at photon 

energies around 100 keV. This over-response is due primarily to the silver (Ag) and bromine (Br) in 

the film emulsions. The response of the newer TLD badges provided a much better match to the 

Hp(10) response in the soft tissues of the body due to the lower atomic numbers of the lithium (Li) and 

fluorine (F) in the TLD chips (Horowitz 1984; Cameron, Sunthanalingham, and Kennedy 1968). The 

two-element TLD badges used at Y-12 from 1980 to 1989 had LiF chips covered by an “open window” 

and an aluminum filter for beta-photon discrimination (McLendon 1980a) and the multi-element TLD 

badges adopted for use at Y-12 in 1989 had four LiF chips covered by an “open window”, a plastic 

filter, a copper filter, and a hemispherical Teflon button (Y-12 Plant 1988b; BWXT Y-12 2001). The 

photon doses were determined primarily from the readings of the LiF chips covered by the aluminum 

filter in the two-element TLD badge and the hemispherical Teflon button in the multi-element TLD 

Badge. A value for “average background radiation” of 0.75 mrem per week from photons was 

determined for the Oak Ridge area by storing a total of 1,680 TLD dosimeters in 70 houses for up to 

one year (Sonder and Ahmed 1991). The distribution of results indicated a rather large variation in 

background among houses, with a few locations exhibiting background double the average. It was 

suggested that the results from the high background houses be ignored in determining values of the 

MDL to be used in routine personnel dosimetry monitoring. 

10|-2 10|-1 10|0

Photon Energy (MeV)

0.1

1

10

Relative Response

Hp(10)

Two-Element Film Badge

Multi-Element Film Badge

Figure 6.3.2.1-1. Comparison of Hp(10) from a broad beam 

of normally incident photons (ICRP 1996) with the energy 

responses for a sensitive DuPont 502 emulsion in a MED 

two-element film badge (Pardue, Goldstein, and Wollan 

1944) and a sensitive DuPont 555 emulsion in an ORNL 

multi-element film badge (Thornton, Davis, and Gupton 

1961). 

The IARC has conducted a recent dosimeter intercomparison study of ten commonly-used historical 

dosimetry systems from around the world (Thierry-Chef et al. 2002). Three of the dosimeter designs 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 16 of 47 

were from the United States. These included a two-element film dosimeter previously used at 

Hanford (identified as US-2), a multi-element film dosimeter previously used at Hanford (identified as 

US-8), and the Panasonic TLD dosimeter currently used at the Savannah River Site (identified as US22). The IARC study considered that exposure to workers could be characterized as a combination of 

anterior-posterior (A-P), rotational, and isotropic irradiation geometries. Dosimeter responses for 

these various irradiation geometries were investigated using two different phantoms to represent the 

torso of the body. The first phantom was a water-filled slab phantom with polymethyl methacrylate 

walls, an overall width of 30 cm, an overall height of 30 cm, and an overall depth of 15 cm. This 

phantom is widely used for dosimeter calibration and performance testing by the International 

Standards Organization (ISO). The second phantom was an anthropomorphic Alderson Rando 

Phantom. This realistic man-type phantom is constructed using a natural human skeleton cast inside 

material that has a tissue equivalent composition. The results of this IARC study, for the U.S. 

dosimeters only, are presented in Table 6.3.2.1-1. As noted previously, the multi-element film badge 

was used at Y-12 in essentially the same manner as the two-element film badge. It should also be 

noted that the two-element film dosimeter can significantly overestimate Hp(10) at the lower photon 

energies of 118 keV and 208 keV. 

Table 6.3.2.1-1. IARC study results for US beta/photon dosimeters.

118 keV 208 keV 662 keV 

Geometry Phantom Meana

 SD/Mean Meana SD/Mean Meana SD/Mean 

US-2 (Two-element film dosimeter)

A-P Slab 3.0 2.1 1.3 1.0 1.0 0.8 

A-P Anthropomorphic 3.0 4.2 1.2 1.9 1.0 1.8 

Rotational Anthropomorphic 2.2 2.0 1.4 3.0 1.2 3.2 

Isotropic Anthropomorphic 1.5 4.4 1.1 1.6 1.0 2.7 

US-8 (Multi-element film dosimeter)

A-P Slab 1.0 1.5 1.0 0.8 0.8 1.7 

A-P Anthropomorphic 0.8 9.5 0.9 6.0 0.8 1.8 

Rotational Anthropomorphic 1.2 1.9 1.2 17 1.1 1.8 

Isotropic Anthropomorphic 1.0 3.0 1.2 9.0 1.0 2.3 

US-22 (Multi-chip TLD dosimeter)

A-P Slab 0.9 4.4 0.9 3.9 0.9 3.5 

A-P Anthropomorphic 0.8 3.1 0.9 2.1 0.9 3.9 

Rotational Anthropomorphic 1.1 3.1 1.2 1.5 1.0 4.1 

Isotropic Anthropomorphic 0.9 0.3 1.0 2.5 0.9 1.6 

a. Ratio of recorded dose to Hp(10). 

6.3.2.2 Neutron Dosimeters 

The two general types of neutron dosimeters that have been used at Y-12 differ significantly in their 

response to neutrons of different energies as illustrated in Figure 6.3.2.2-1 (IAEA 1990). An NTA 

emulsion was included in the same holder used for the Y-12 beta/gamma dosimeter until 1980. 

Between 1980 and 1989, there is a serious gap in the neutron dosimetry information for Y-12. It is 

known that Y-12 had become increasingly dependent over the years on ORNL to process the NTA 

films because of the small numbers of neutron-exposed workers at Y-12. Thus, the neutron dosimetry 

at Y-12 is assumed to be the same as that at ORNL from 1980 to 1989. During this period, workers at 

both ORNL and Y-12 were provided with a two-element TLD dosimeter for beta-particle and photon 

dosimetry. Those ORNL and Y-12 workers who were exposed to neutrons were provided with a 

separate neutron dosimeter. This neutron dosimeter contained both an NTA film for measurement of 

the fast-neutron dose and a TLND for measurement of the neutron dose from lower energy neutrons 

(Gupton 1978; Berger and Lane 1985). From 1980 to 1985, the neutron doses to Y-12 workers were 

determined at ORNL using both the NTA and TLND dosimeters as discussed by Gupton (1978). 

From 1986 to 1989, they were determined at ORNL using only the TLNDs (Berger and Lane 1985). 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 17 of 47 

Since 1989, the neutron dose to Y-12 workers has been measured using a separate albedo-type 

TLND that is worn on the belt to keep it in close contact with the worker’s body (Gunter 1994; BWXT 

Y-12 2001). In general, the response of the NTA film decreases with decreasing neutron energy that 

are greater than a threshold energy estimated to be about 500 keV (IAEA 1990), and the TLND 

response increases with decreasing neutron energy as illustrated in Figure 6.3.2.2-1. Results reported 

at the first AEC Neutron Dosimetry Workshop in 1969 indicated that laboratory dose measurements 

made with NTA film were about one-half to one-fourth of those measured with other methods 

including the TLND (Vallario, Hankins, and Unruh 1969). The response of both dosimeters is highly 

dependent upon the neutron energy spectra, and both dosimeter types require matching the 

laboratory calibration neutron spectra to the workplace neutron spectra for reliable results. 

10|-7 10|-6 10|-5 10|-4 10|-3 10|-2 10|-1 10|0 10|1

Neutron Energy (MeV)

10|0

10|1

10|2

10|3

10|4

Relative Response

Hp(10)

NTA Film

6

LiF with Cd Shield

Figure 6.3.2.2-1. Comparison of Hp(10) from normally 

incident neutrons (IAEA 2001) to the energy responses of a 

nuclear track type A (NTA) film and a neutron albedo 

dosimeter containing a neutron thermoluminescent 

dosimeter (TLND) chip made of lithium-6 fluoride and 

shielded by cadmium (IAEA 1990, 2001). 

6.3.3 Y-12 Dosimeter Calibration Procedures

Potential error in recorded dose is dependent on the methodology used to calibrate the dosimeters 

and the extent of the similarity between the radiation fields used for calibration and those encountered 

in the workplace. The potential error is much greater for dosimeters with significant variations in 

response such as film dosimeters for low energy photon radiation and both the NTA and TLND 

dosimeters for neutron radiation. 

6.3.3.1 Beta/Photon Dosimeters 

The Y-12 dosimeters were originally calibrated primarily using beta-particles from a natural uranium 

slab and photons from a 226Ra source (Souleyrette 2003). The dosimeters were exposed facedown 

on the uranium slab and free in air (i.e., no phantom) facing the 226Ra source for pre-selected times to 

produce beta-particle and photon doses normally encountered in the workplace. This practice was 

similar to that used at other AEC sites. Several common sources of expected laboratory bias are 

discussed in Table 6.3.3.1-1 for personnel beta/photon dosimeter calibration based on comparison of 

the recorded dose with Hp(10). 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 18 of 47 

Table 6.3.3.1-1. Common sources of laboratory bias in the calibration parameters for beta/photon 

dosimeters.a

Parameter Historical description Anticipated laboratory biasb

Free-in-air 

calibration 

In the 1980s, Y-12 began exposing 

calibration dosimeters on a phantom to 

simulate a worker’s body. The previous 

calibrations do not include response from 

backscattered radiation. 

Recorded dose of record is too high; however, 

the effect of backscattered radiation from 

worker’s body is highly dependent on the 

dosimeter design. 

Radiation 

quantity 

Photon dose quantities that were used to 

calibrate Y-12 beta/photon dosimeters 

have varied historically. The exposure

from photons was used for many years. 

Because of the higher energy, 226Ra and 60Co 

gamma radiation used to calibrate dosimeters 

at Y-12, this caused only a slight (about 3%) 

under-response in the recorded dose. 

Depth of tissue 

dose 

Historically, Y-12 used a selected depth 

of 1 cm (i.e., the depth of the testes) to 

estimate the deep dose. 

No significant effect because the Y-12 

dosimeter designs had filtration density 

thicknesses of about 1,000 mg cm-2 that would 

relate to the 1-cm depth in tissue. 

Angular 

response 

Y-12 dosimeters are calibrated using 

anterior-posterior (A-P) laboratory 

irradiation. 

Recorded dose of record is likely too low 

because the dosimeter response is lower at 

non A-P angles. The effect is highly 

dependent on radiation type and energy. 

Environmental 

stability 

Y-12 film and TLD dosimeters are 

subject of signal fade with time, heat, 

humidity, light, etc. 

Recorded dose of record is likely too low; 

however, this depends strongly upon when the 

calibration dosimeters are irradiated during the 

dosimeter exchange cycle. Mid-cycle 

calibration minimizes the effects. 

a. Judgment based on Y-12 dosimeter response characteristics. 

b. Recorded dose compared to Hp(10). 

6.3.3.2 Neutron Dosimeters 

A good account of the historical aspects of the calibration of the Y-12 neutron dosimeters is not 

available. It is known, however, that the NTA films were originally calibrated using a Po-Be neutron 

source (Struxness 1949b, 1952a) and later an Am-Be neutron source (McLendon 1963; McRee, 

West, and McLendon 1965). The dosimeters containing the NTA films were exposed free in air (i.e., 

no phantom) to neutrons from the Po-Be and Am-Be sources for pre-selected times to produce 

neutron doses normally encountered in the workplace. Some information on the calibration of Y-12 

neutron dosimeters containing TLNDs can be found in Berger and Lane (1985) and BXWT Y-12 

(2001). Several common sources of expected laboratory bias are discussed in Table 6.3.3.2-1 for 

personnel neutron dosimeters based on comparison of the recorded dose with Hp(10). 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 19 of 47 

Table 6.3.3.2-1. Common sources of laboratory bias in the calibration parameters for neutron 

dosimeters.a

Parameter Historical description Anticipated laboratory biasb

Source energy 

spectrum 

In 1980, Y-12 began using dosimeters that 

were calibrated on a phantom to simulate a 

worker’s body and neutron spectra that were 

degraded to better represent the workplace. 

The previous calibrations did not include 

response from backscattered radiation and 

the neutron spectra were not degraded. 

The delivered dose was uncertain as 

noted in Section 6.3.2.2 of this report. 

Radiation 

quantity 

Neutron dose quantities that were used to 

calibrate Y-12 neutron dosimeters have 

varied historically. The first collision dose for 

fast neutrons and a quality factor of 10 was 

used for many years. 

The effects of the respective neutron dose 

quantities used to calibrate Y-12 

dosimeters is uncertain and could be 

evaluated in comparison to the Hp(10) 

dose used in DOELAP performance 

testing. 

Angular 

response 

Y-12 dosimeters are calibrated using 

anterior-posterior (A-P) laboratory irradiation. 

Recorded dose of record is likely too low 

because the dosimeter response is lower 

at non A-P angles. The effect is highly 

dependent on neutron energy. 

Environmental 

stability 

Y-12 NTA film and TLND dosimeters are 

subject of signal fade with time, heat, 

humidity, light, etc. 

Recorded dose of record is likely too low; 

however, this depends strongly upon 

when the calibration dosimeters are 

irradiated during the dosimeter exchange 

cycle. Mid-cycle calibration minimizes the 

effects. 

a. Judgment based on Y-12 dosimeter response characteristics. 

b. Recorded dose compared to Hp(10). 

6.3.4 Y-12 Workplace Radiation Fields

The main workplace radiation fields at Y-12 are due to processes involving either EU (235U) or 

depleted uranium (DU) (238U). Some other workplace radiation fields involve industrial radiation 

generating equipment (x-rays and electron accelerators) and isotopic gamma-ray and neutron 

sources for testing purposes (60Co and 252Cf). The current Y-12 External Dosimetry Technical Basis 

document provides a discussion of the radiation fields due to different processes and primary nuclides 

in the Y-12 workplace (BWXT Y-12 2001). 

6.3.4.1 Workplace Beta/Photon Dosimeter Response 

Radiation (beta/photon) fields characteristic of the Y-12 facilities can be generally defined based on 

historical information as presented in Table 6.3.4.1-1. Because Y-12 is a nuclear weapons fabrication 

and disassembly facility, the most common materials are EU (235U) and DU (238U). Both 235U and 238U 

are primarily alpha-particle emitters. However, 235U does emit a 185-keV photon in 54% of its decays. 

Most of the external dose from 238U comes from its short-lived 234Th, 234mPa, and 234Pa decay 

products. From an external dose standpoint, the most significant radiations emitted by these decay 

products of 238U are: (1) the 2.29-MeV beta particle from 234mPa, and (2) the photons emitted by 234Pa 

with energies as large as 1.962 MeV. The various Y-12 dosimeters have filtration of about 1,000 mg 

cm-2 (i.e., nearly equivalent to 1-cm depth in tissue) for those regions of the dosimeter used to 

measure the whole-body dose. The response to beta radiation in Y-12 workplaces is limited because 

beta radiation usually cannot penetrate this much filtration. 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 20 of 47 

Table 6.3.4.1-1. Selection of beta and photon radiation energies and percentages for Y-12 site 

processes. 

Operations

Y-12 site processes Building Begin End

Radiation 

type

Energy 

selection Percent

9203 1947 1951 

9206a 1947 1959 

9211 1947 1959 

Enriched uranium product recovery 

and salvage operations 

9201-1 1952 1963 

Beta 

Photon 

> 15 keV 

30-250 keV 

100% 

100% 

9202 1947 1995 

9206a 1947 1995 

Uranium chemical operations and 

weapon production operations 

9212b

 1949 Ongoing 

Beta 

Photon 

> 15 keV 

30-250 keV 

100% 

100% 

Special nuclear material receiving 

and storage 

9720-5 1949 Ongoing Photon 30-250 keV 100% 

9201-5 1949 Ongoing 

9204-4 1949 Ongoing 

Uranium forming and machining for 

weapon component operations 

9215 1950 Ongoing 

Beta 

Photon 

> 15 keV 

30-250 keV 

100% 

100% 

9201-5 1949 Ongoing 

9204-4 1949 Ongoing 

9766 1949 ? 

Depleted uranium process 

operations 

9998 1949 Ongoing 

Beta 

Photon 

> 15 keV 

30-250 keV 

> 250 keV 

100% 

50% 

50% 

Final weapon component 9204-2 1952 Ongoing 

assembly operations 9204-2E 1952 Ongoing 

Beta 

Photon 

> 15 keV 

30-250 keV 

100% 

100% 

ORNL 86-inch cyclotron 9201-2 1950 ? Photon 30-250 keV 

>250 keV 

50% 

50% 

Chemical assay and mass 

spectrometry laboratories 

9203 1947 Ongoing Photon 

Radiographic laboratory 9201-1 1947 Ongoing Photon 

Calibration laboratory 9983 1949 Ongoing Photon 

Weapon component assay 

laboratory 

9995 1952 Ongoing Photon 

Nondestructive assay laboratory 9720-5 1980 Ongoing Photon 

Specific to 

radiation 

source. 

Photon 

default 

values: 

30-250 keV 

>250 keV 

50% 

50% 

West End waste treatment facility 

9616-7 1984 Ongoing Beta 

Photon 

> 15 keV 

30-250 keV 

>250 keV 

100% 

50% 

50% 

a. Building 9206 Complex includes Buildings 9768, 9720-17, 9409-17, 9510-2, 9767-2, and the east and west tank farm pits. 

b. Building 9212 Complex includes Buildings 9809, 9812, 9818, 9815, and 9980. 

The largest workplace exposures at Y-12 have historically occurred in the DU process areas 

(Struxness 1952b; Henderson 1991). During casting operations, the decay products of 238U float to 

the top surface of the molten metal and remain as surface residues. These surface residues result in 

an increased exposure potential because of the high beta and photon energies associated with the 

234Pa nuclide. The 234Pa nuclide emits a number of high-energy photons and has a specific activity 

that is approximately 2 x 1015 times larger than the specific activity of its 238U parent (Henderson 

1991). For 234Pa, the percentages of photons with energies of 30 – 250 keV and 250 keV or more are 

about 7 and 93%, respectively, and for 238U in equilibrium with its short-lived 234Th, 234mPa, and 234Pa, 

the percentages of photons with energies of 30 – 250 keV and 250 keV or more are about 82 and 

18%, respectively. Thus, an artificially high percentage of photons with energies greater than 250 keV 

was assumed in Table 6.3.4.1-1 for the normal and depleted uranium process areas. This produces 

doses that are claimant favorable because of the increased exposure potential to high energy photons 

from the short-lived 234Pa decay product of 238U. 

Typical beta/photon personnel dosimeter parameters important to Hp(10) performance in the 

workplace are summarized in Table 6.3.4.1-2. 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 21 of 47 

Table 6.3.4.1-2. Typical workplace beta/gamma dosimeter Hp(10) performance.a

Parameter Description Potential workplace biasb

Exposure 

geometry 

Y-12 dosimeter system calibrated using A-P 

laboratory irradiations. 

Recorded dose of record likely too low 

since dosimeter response is lower at angles 

other than A-P. Effect is highly dependent 

upon radiation type and energy. 

Energy response 

Y-12 film “deep dose” response is too low 

for photon radiation less than 100 keV and 

too high for photon radiation greater than 

100 keV. 

Bias is recorded dose depends upon the 

photon energy in the workplace. 

Highly divergent 

fields 

Dosimeter worn at collar may underestimate the deep dose at the waist. Recorded dose of record may be too low for 

workers performing waist-level uranium 

handling jobs. 

Mixed fields Y-12 dosimeters respond to both beta and 

photon radiation. 

Filtration of about 1,000 mg cm-2 over 

dosimeter region used to measure deep 

dose will minimize dosimeter response to 

beta radiation. 

Missed dose Doses less than Minimum Detection Level 

(MDL) recorded as zero dose. 

Recorded dose of record likely too low. The 

impact of missed dose is greatest in the 

earlier years because of frequency 

dosimeters exchange and film dosimeter 

with higher MDLs. 

Environmental 

effects 

Workplace environment (heat, humidity, 

etc.) fades the dosimeter signal. 

Recorded dose of record is likely too low. 

a. Judgment based on Y-12 dosimeter response characteristics and workplace radiation fields. 

b. Recorded dose compared to Hp(10). 

One serious problem with the workplace response of the Y-12 beta/gamma dosimeters involves 

workers who perform waist-level uranium handling jobs in the DU process areas (Henderson 1991). 

A personnel dosimeter worn at the collar may underestimate the Hp(10) dose at the waist by rather 

significant factors. It is now a practice to instruct these workers to wear the dosimeters at the waist, 

but many workers may have worn their dosimeters on the collar in the past. Hence, for all workers 

performing waist-level uranium handling jobs, it is recommended that the recorded dose before 1991 

should be multiplied by 1.34 for reasons discussed in more detail by Henderson (1991). To determine 

when to make such adjustments, the dose reconstructor must depend on information about routine 

duties and work locations that are contained in the Computer Assisted Telephone Interview (CATI) file 

for a claimant. 

6.3.4.2 Workplace Neutron Dosimeter Response 

Three main facilities at Y-12 with a potential for neutron exposure are: (1) the Calibration Laboratory 

in Building 9983, (2) the EU Storage Area in Building 9212, and (3) the Nondestructive Assay 

Laboratory in Building 9720-5. The following sections discuss the neutron exposure spectra and 

neutron-to-photon dose ratios in these three areas using data from recent measurements made the by 

Pacific Northwest Laboratory (PNL 1990; McMahan 1991; BWXT Y-12 2001). 

6.3.4.2.1 Calibration Laboratory in Building 9983 

The Calibration Laboratory has a highly shielded room used for storage of both photon and neutron 

sources. The walls of the room are all 3-inch steel, with a high-density concrete floor into which 

several source storage pits are sunk. The types of neutron sources stored in this room include twelve 

2 to 4 Ci americium boron (Am-B) sources, several americium lithium (Am-Li) sources, and several 

americium beryllium (Am-Be) sources. At the time of the PNL measurements, the neutron sources 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 22 of 47 

were stored in the room in shielded containers. The neutron shielding of the containers was either 

paraffin or high-density plastic, depending on the container. Several sources were stored in 

containers inside a steel safe, others were in their containers on the floor of the room, and still others 

were in the storage pits below floor level. The neutron sources have not been used for routine 

calibration purposes since the early 1970s, when ORNL began calibrating all Y-12 neutron detection 

and survey instruments. Workers at Y-12 do access the source storage area for other purposes and 

the PNL measurements were made outside the door to the source storage room to determine 

appropriate calibration factors for a worker’s TLND. 

6.3.4.2.1.1 Neutron Energy Spectrum 

The PNL measurements of the neutron energy spectrum at a distance of 18 inches from the door to 

the source storage room are shown by the solid line in Figure 6.3.4.2-1. It should be noted that the 

PNL measurement data were provided as dose equivalent rates (PNL 1990); however, a one-hour 

exposure was assumed here in order to show the results of the PNL measurements as dose 

equivalent. The fluence-to-dose equivalent conversion factors and neutron quality factors used in the 

PNL measurements are similar to those from National Council on Radiation Protection and 

measurements (NCRP) Report 38 (1971) and International Commission on Radiological Protection 

(ICRP) Publication 21 (1973). A comparison of fluence-to-dose equivalent conversion factors from 

NCRP Report 38, ICRP Publication 21, and several other commonly used information sources on 

10|-7 10|-6 10|-5 10|-4 10|-3 10|-2 10|-1 10|0 10|1 10|2

Neutron Energy (MeV)

10|-4

10|-3

10|-2

10|-1

10|0

10|1

Dose Equivlanet (mrem)

0.255 mrem 0.0304 mrem

2.38 mrem 2.01 mrem

Figure 6.3.4.2-1. Results of PNL neutron spectrum measurements in 

Calibration Laboratory of Building 9983 are shown by the solid line in the 

graph and the dashed line shows the PNL measurement results divided into 

the four neutron energy groups used in the dose reconstruction for the Y-12 

workers. 

fluence-to-dose equivalent conversion factors and neutron quality factors can be found in a report by 

Sims and Killough (1983). The dashed line in Figure 6.3.4.2-1 shows the dose equivalent from the 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 23 of 47 

PNL measurements divided into the four energy groups discussed in NIOSH 2002, and the dose 

fractions in each of these four energy groups are provided in Table 6.3.4.2-1. Although PNL 

measured some dose from lower (< 10 keV) and intermediate energy (10-100 keV) neutrons, the 

contribution to the total dose was only about 6%. The Radiation Effectiveness Factor (REF) used in 

the Interactive Radio-Epidemiological Program (IREP) to calculate the Probability of Causation (PC) 

for these two neutron energy groups is less than the fast neutron energy group (or so-called fission 

neutron energy group) from 0.1-2 MeV (Kocher, Apostoaei, and Hoffman 2002). As a result, 

combining the lower and intermediate energy groups into the fast neutron group from 0.1-2 MeV is a 

reasonable and claimant-favorable simplification of the neutron dose calculation. 

Table 6.3.4.2-1. Dose fractions for Y-12 calibration laboratory. 

Neutron energy group Near source storage safe

< 10 keV 0.055 

10-100 keV 0.007 

0.1-2 MeV 0.509 

2-14 MeV 0.429 

Claimant-favorable dose fractions

0.1-2 MeV 0.57 

2-14 MeV 0.43 

6.3.4.2.1.2 Neutron-to-Photon Dose Ratio 

The neutron-to-photon dose ratio from the recent PNL measurements was approximately 8:1. For 

workers in the Calibration Laboratory, no other data on neutron-to-photon dose ratios have been 

found to use to estimate the missing neutron dose from measurements made with film dosimeters. 

However, the recent PNL studies indicate that more that 90% of the neutron dose is above the 500-

keV threshold of the NTA films. Also, neutron dose measurements for Calibration Laboratory workers 

with NTA film dosimeters are expected to be reasonably accurate to within parameters discussed in 

NIOSH 2002. 

6.3.4.2.2 Enriched Uranium Storage Area in Building 9212 

Building 9212 contains a secure storage area for containers of enriched uranium fluoride (UF4) and 

uranium trioxide (UO3). Neutrons are produced by alpha particle reactions with the nucleus of the 

fluorine and oxygen atoms of the UF4 and UO3, respectively. Containers of these materials are 

placed on a rack of shelves and arranged in a matrix that is critically safe. The containers are spaced 

approximately 2 to 2.5 feet apart on a shelf, and can be placed one deep per shelf. There are four 

shelves per rack and 24 inches between shelves. The PNL measurements were made 39 inches 

above the floor and 24 inches from the shelf at a location near the center of a rack that was filled with 

20 containers of UF4. 

6.3.4.2.2.1 Neutron Energy Spectrum 

The PNL measurements of the neutron energy spectrum near the center of the UF4 storage rack are 

shown by the solid line in Figure 6.3.4.2-2, and the dose fractions for the neutron energy groups 

shown by the dashed line in this figure are provided in Table 6.3.4.2-2. The dose fraction for the 

lower (<10 keV) and intermediate (10-100 keV) energy neutron groups were less than 2% of the total 

dose from these PNL measurements. As before, combining the lower and intermediate energy 

groups into the fast neutron group from 0.1-2 MeV is a reasonable and claimant favorable 

simplification of the neutron dose calculation. 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 24 of 47 

10|-7 10|-6 10|-5 10|-4 10|-3 10|-2 10|-1 10|0 10|1 10|2

Neutron Energy (MeV)

10|-4

10|-3

10|-2

10|-1

10|0

10|1

Dose Equivlanet (mrem)

0.0193 mrem

 1.29 mrem

0.00381 mrem

0.340 mrem

Figure 6.3.4.2-2. Results of PNL neutron spectrum measurements in 

the Enriched Uranium Storage Area of Building 9212 are shown as the 

solid line in the graph. The dashed line shows the PNL measurement 

results divided into the four neutron energy groups used in the dose 

reconstruction for the Y-12 workers. 

Table 6.3.4.2-2. Neutron dose fractions for Y-12 Enriched 

Uranium Storage Area. 

Neutron energy group Near storage rack

< 10 keV 0.012 

10-100 keV 0.002 

0.1-2 MeV 0.781 

2-14 MeV 0.205 

Claimant-favorable dose fractions 

0.1-2 MeV 0.79 

2-14 MeV 0.21 

6.3.4.2.2.2 Neutron-to-Photon Dose Ratio 

The neutron-to-photon dose ratio from the recent PNL measurements was approximately 1:1. For 

workers in the Enriched Uranium Storage Facility, no other data on neutron-to-photon dose ratios in 

Building 9212 have been found to use to estimate the missing dose from earlier measurements made 

with film dosimeters. However, the recent PNL studies indicate more than 95% of the neutron dose is 

above the 500-keV threshold of the NTA films. Also, the neutron dose measurements for workers in 

the Enriched Uranium Storage Area of Building 9212 with NTA film dosimeters are expected to be 

reasonably accurate to within parameters discussed in NIOSH 2002. 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 25 of 47 

6.3.4.2.3 Nondestructive Assay Laboratory in Building 9720-5 

The Nondestructive Assay Laboratory in Building 9720-5 is used for recovery of highly enriched 

uranium (HEU) from manufacturing wastes (Hogue and Smith 1984). The laboratory contains 

instruments for gamma scanning and neutron interrogation of containers of solid wastes, gamma 

analysis of solution samples, and measurements of solution density. Because measurements of the 

neutron spectrum were made previously for a 252Cf fission-neutron source at ORNL’s Radiation 

Calibration Laboratory (RADCAL), it was not necessary to make additional neutron measurements to 

characterize the workplace radiation fields near the 252Cf neutron source at the Nondestructive Assay 

Laboratory at Y-12. The neutron measurements at the RADCAL facility were made at a distance of 

39 inches from the bare 252Cf source and a height of approximately 39 inches above the floor. 

6.3.4.2.3.1 Neutron Energy Spectrum 

The results of the PNL neutron spectrum measurements made at 39 inches from the 252Cf fissionneutron source at ORNL’s RADCAL facility are shown by the solid lines in Figure 6.3.4.2-3. The dose 

fractions for the neutron energy groups shown by the dashed lines in this figure are provided in Table 

6.3.4.2-3. The dose fractions for the lower (<10 keV) and intermediate (10-100 keV) energy neutron 

groups were less than 1% of the total dose from these PNL measurements. Thus, combining the 

lower and intermediate energy groups into the fast neutron group of 0.1 to 2 MeV is a reasonable and 

claimant favorable simplification of the neutron dose calculation. 

10|-7 10|-6 10|-5 10|-4 10|-3 10|-2 10|-1 10|0 10|1 10|2

Neutron Energy (MeV)

10|-7

10|-6

10|-5

10|-4

10|-3

10|-2

10|-1

Dose Equivlanet (mrem)

9.75e-5 mrem 1.21e-4 mrem

7.59e-3 mrem

2.61e-2 mrem

Figure 6.3.4.2-3. Results of PNL neutron spectrum measurements made 

at 1 meter from a bare 252Cf fission neutron source are shown by the solid 

line in the graph and the dash line shows the PNL measurement results 

divided into the four neutron energy groups used in the dose 

reconstruction for Y-12 workers at the Nondestructive Analysis Laboratory 

in Building 9720-5. 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 26 of 47 

Table 6.3.4.2-3. Dose fractions for Y-12 nondestructive analysis 

laboratory. 

Neutron energy group Near unshielded Cf-252 source

< 10 keV 0.003 

10-100 keV 0.004 

0.1-2 MeV 0.224 

2-14 MeV 0.769 

Claimant-favorable dose fractions

0.1-2 MeV 0.23 

2-14 MeV 0.77 

6.3.4.2.3.2 Neutron-to-Photon Dose Ratio 

The neutron-to-photon dose ratio from the recent PNL measurements was approximately 25:1. For 

workers in the Nondestructive Assay Laboratory in Building 9212, no other data on neutron-to-dose 

ratios have been found to use to estimate missing dose from earlier measurements made with film 

dosimeters. However, the recent PNL studies indicate that more than 97% of the neutron dose is 

above the 500-keV threshold of the NTA films. Also, the neutron dose measurements for workers in 

the Nondestructive Assay Laboratory with NTA film dosimeters are expected to be reasonably 

accurate to within the parameters discussed in NIOSH (2002). 

6.3.4.3 Typical Workplace Neutron Dosimeter Hp(10) Performance 

Typical neutron personnel dosimeter parameters important to Hp(10) performance in the workplace 

are summarized in Table 6.3.4.3-1. The most important parameter related to Hp(10) performance of 

the neutron dosimeters is the difference between calibration and workplace neutron energy spectra. 

Measurements made by PNL in the late 1980s and early 1990s, could be used to correct the 

response of the Y-12 TLND dosimeters to workplace neutron energy spectra in several Y-12 areas. 

These measurements were discussed in the previous section of this report and the results of the 

corrected TLND workplace measurements over a 12-year period starting in 1990 are shown in Table 

6.3.4.3-2. These data illustrate the low potential for routine exposure to neutrons at Y-12 both now 

and in the past. 

Table 6.3.4.3-1. Typical workplace neutron dosimeter Hp(10) performance.a

Parameter Description Potential workplace biasb

Workplace neutron 

energy spectra 

NTA dosimeter response decreases and 

TLND response increases with 

decreasing neutron energy 

Depends upon workplace neutron 

spectra. NTA recorded dose of record 

likely too low because of high 500-keV 

threshold for detection of neutrons. 

Exposure geometry 

NTA dosimeter response increases with 

increasing exposure angle and TLND 

response decreases with 

increasing exposure angle. 

NTA recorded dose likely too high since 

dosimeter response is higher at angles 

other than A-P. TLD recorded dose is 

lower at angles other than A-P. Effect is 

highly dependent on neutron energy. 

Missed dose 

Doses less than Minimum Detection Limit 

(MDL) recorded as zero dose. 

Recorded dose of record is likely too low. 

The impact of missed dose is greatest in 

earlier years because of the higher MDLs 

of the neutron dosimeters. 

Environmental effects Workplace environment (heat, humidity, 

etc.) fades the dosimeter signal. 

Recorded dose of record is likely too low. 

a. Judgment based on Y-12 dosimeter response characteristics. 

b. Recorded dose compared to Hp(10). 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 27 of 47 

Table 6.3.4.3-2. Number of neutron monitored 

workers, cumulative neutron dose, and average 

neutron dose to Y-12 workers for 12-year period 

following introduction of the four-element TLND 

dosimeter in 1989. 

Year

Neutron 

monitored 

workers

Cumulative 

neutron dose 

(mrem)

Average 

neutron dose 

(mrem)

1990 82 1085 13.2 

1991 64 463 7.2 

1992 86 200 2.3 

1993 215 343 1.6 

1994 301 1289 4.3 

1995 165 116 0.7 

1996 203 470 2.3 

1997 38 10 0.3 

1998 47 57 1.2 

1999 141 121 0.9 

2000 49 35 0.7 

2001 73 55 1.3 

The recent PNL measurements also indicate that the past NTA film dosimeters worked reasonably 

well in the Y-12 workplace because the Ra-Be and Po-Be neutron spectra used to calibrate them 

were reasonably well matched to the workplace neutron spectra. These measurements suggest that 

the NTA film dosimeters missed less than 10% of the neutron dose equivalent at the Calibration 

Laboratory of Building 9983 and less than 5% of the neutron dose equivalent at the Enriched Uranium 

Storage Area of Building 9212 and the Nondestructive Assay Laboratory of Building 9720-5. It must 

be noted that there are a lot of recorded zeros in the neutron dose data for Y-12 workers for two 

reasons: (1) a worker’s NTA film was not developed and read, or (2) a worker’s neutron dose 

equivalent was less than the MDL for the NTA film. 

6.4 ADJUSTMENTS OF RECORDED DOSE 

Adjustments to the Y-12 recorded doses are necessary to arrive at a claimant-favorable dose, 

considering the uncertainty associated primarily with the complex workplace radiation fields and 

exposure geometries. 

6.4.1 Photon Dose Adjustments

The average and maximum deep photon dose to Y-12 workers for the 10-year period from 1978 to 

1987 is shown in Figure 6.4.1-1 (Y-12 Plant 1978, 1980, 1981, 1982b, 1983, 1984, 1985, 1986, 1987, 

1988a). The UCC-ND policy at that time was to limit the maximum deep photon dose to workers to 

500 mrem or less per quarter and 2,000 mrem or less per year. The average deep dose from photons 

to all Y-12 workers was approximately 20 mrem from 1978 to 1987. This time period covers the 

change from film dosimeters to TLD dosimeters in 1980. It should be noted that no abrupt change 

occurred in the deep penetrating data for photon dose in 1980. Hence, the recorded doses for the 

photon deep dose from both film and TLD dosimeters appear to be in very close agreement and no 

adjustments are deemed necessary to the recorded deep photon doses for most Y-12 workers. 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 28 of 47 

1978 1979 1980 1981 1982 1983 1984 1985 1986 1987

Year

10|0

10|1

10|2

10|3

10|4

10|5

Annual Dose Equivalent (mrem)

Maximum photon deep dose

Average photon deep dose

Figure 6.4.1-1. Maximum and average photon deep dose to 

Y-12 workers for the 10-year period from 1978 to 1987. 

There is one group of Y-12 workers for which an adjustment in the recorded photon dose is 

recommended (see Section 6.3.4.1). These workers performed waist-level handing jobs in DU 

process areas (see Table 6.3.4.1-1). Examples of waist-level handling jobs are the unloading and 

sorting of DU scrap materials, shearing of larger pieces of scrap materials, cleaning of the scrap 

materials, crucible loading during the melting and casting operations, and materials sampling 

(Henderson 1991). It is now a practice to instruct workers performing these operations to wear their 

beta/photon dosimeters at the waist, but many of these workers may have worn their dosimeters at 

the collar in the past. The photon dose correction summarized in Table 6.4.1-1 is necessary to 

calculate an adjusted photon dose that is claimant favorable prior to 1991. For the years 1991 to the 

present, no correction is needed because the recorded photon dose is Hp(10) equivalent. 

Table 6.4.1-1. Adjustments to reported Y-12 deep photon dose. 

Parameter Description

Period Prior to 1/1/1991 

Dosimeters All beta/photon dosimeters 

Facilities Depleted uranium process operations 

Workers Waist-level metal handling operators 

Adjustment to recorded dose Multiply reported deep photon dose by a 

factor of 1.34 to estimate Hp(10) 

6.4.2 Neutron Dose Adjustments

The Y-12 facility incorporated the energy variation of the dose equivalent from neutrons into their 

calibration methodology. Thus, the recorded dose equivalent is a combination of all neutron energies. 

To calculate the probability of causation, the recorded neutron dose must be separated into neutron 

energy groups as discussed in Section 6.3.4.2 and then converted into ICRP Publication 60 (1990) 

methodology. 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 29 of 47 

6.4.2.1 Neutron Weighting Factors 

An adjustment to the neutron dose is necessary to account for the change in neutron quality factors 

between historical and current scientific guidance as discussed by NIOSH (2002). At Y-12, the 

TLNDs were calibrated using PNL measurements based on fluence-to-dose conversion factors and 

quality factors similar to those from ICRP Publication 21 (1973) and NCRP Report 38 (NCRP 1971). 

These quality factors are point-wise data because they were calculated for a broad-parallel beam of 

monoenergetic neutrons incident on a 30-cm diameter cylinder of tissue representing the torso. The 

NCRP Report 38 quality factors are compared in Figure 6.4.2.1-1 with those used in the PNL 

measurements at Y-12. In order to convert from NCRP 38 quality factors to ICRP Publication 60 

radiation weighting factors, a curve was fit that described the quality factors as a function of neutron 

energy. A group average quality factor was then calculated as shown in Figure 6.4.2.1-1 for each of 

the neutron energy groups used to define the radiation weighting factors in ICRP Publication 60 

(1990). A summary of the group averaged NCRP Report 38 quality factors used in the dose 

reconstruction is provided in Table 6.4.2.1-1. This table also compares the group averaged NCRP 38 

quality factors with historical dosimetry guidelines from the First Tripartite Conference at Chalk River 

in 1949 (Fix et al. 1994). 

10|-8 10|-7 10|-6 10|-5 10|-4 10|-3 10|-2 10|-1 10|0 10|1

Neutron Energy (MeV)

0

5

10

15

Neutron Quality Factor

PNL Measurements

NCRP 38: Point-wise data

NCRP38: Group averaged data

Figure 6.4.2.1-1. Comparison of the neutron quality factors used 

in the PNL neutron spectrum measurements and the neutron 

quality factors from NCRP Report 38 (NCRP 1971) shown both as 

point-wise data and grouped averaged data over the four neutron 

energy groups used in the dose reconstruction for the Y-12 

workers. 

6.4.2.2 Neutron Dose Correction Factors 

The average quality factor for the four energy groups that encompass the Y-12 neutron exposures are 

provided in Table 6.4.2.1-1. The neutron dose equivalent correction factor for each of these four 

energy groups, Cf(En), can be calculated by the use of the following equation: 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 30 of 47 

( ) ( )

( ) C E ( ) D E

Q E f n w E f n

avg n

= × R n

where Df(En) is the dose fraction from section 6.3.4.2 for the specific neutron energy group of interest, 

Qavg(En) is the group average NCRP 38 neutron quality factor for that specific energy group, and 

Wr(En) is the ICRP 66 neutron weighting factor for that specific energy group. 

Table 6.4.2.1-1. Neutron quality factor, Q, or weighting factor, wr. 

Neutron energy (MeV) 

Historical dosimetry 

guidelinea

NCRP Report 38 group 

averaged quality 

factorb

ICRP Publication 66 

neutron weighting 

factor 

Thermal 3 

0.5 ev–10 keV 

2.35 5 

10 keV–100 keV 5.38 10 

100 keV–2 MeV 10.49 20 

2 MeV–14 MeV 7.56 10 

14 MeV–60 MeV 

10 

Not applicable 5 

a. First Tripartite Conference at Chalk River in 1949 (Fix et al. 1994). 

b. See Figure 4.6.2-1. 

The neutron dose distributions by energy for the various neutron exposure areas at Y-12 are 

summarized in Table 6.4.2.2-1. By multiplying the recorded neutron dose by the area-specific 

correction factors, the neutron dose equivalent is calculated as follows. Consider security personnel 

who inventory fissile material in the Enriched Uranium Storage Area of Building 9212. Assume that 

the worker receives a recorded annual neutron dose of 100 mrem. The corrected neutron dose is 151 

mrem for neutrons with energies between 0.1-2 MeV, 28 mrem for neutron with energies between 2-

14 MeV, and 179 mrem for neutrons of all energies. These corrections should be applied to both 

measured neutron dose and missed neutron dose. The dose fractions by energy and the associated 

ICRP 60 (1990) correction factors for various neutron exposure areas at Y-12 are summarized in 

Table 6.4.2.2-1. 

Table 6.4.2.2-1. Summary of neutron dose fractions and associated ICRP 60 (1990) correction 

factors for Y-12 facilities. 

Operations

Y-12 facilities Building Begin End Neutron energy

Neutron dose 

fraction

ICRP 60 correction 

factors

Calibration Laboratory 9983 1949 Ongoing 0.1-2 MeV 

2-14 MeV

0.57 

0.43

1.09 

0.57

Enriched Uranium 

Storage Area

9212 1949 Ongoing 0.1-2MeV 

2-14 MeV

0.79 

0.21

1.51 

0.28

Nondestructive 

Analysis Laboratory

9720-5 1980 Ongoing 0.1-2 MeV 

2-14 MeV

0.23 

0.77 

0.44 

1.02 

6.5 MISSED DOSE 

There is undoubtedly missed dose for Y-12 workers. Analysis of the missed dose has been 

separated according to photon and neutron missed dose. The missed photon dose is discussed first 

and then the neutron missed dose. 

6.5.1 Photon Missed Dose

Missed photon dose to Y-12 workers may occur for the following reasons: (1) the worker was not 

monitored before 1961, (2) there is no recorded dose for short periods of time after 1961, and (3) the 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 31 of 47 

worker’s dose during a monitoring period was recorded as zero because the dosimeter response was 

less than the MDL. Before 1961, the policy at Y-12 was to issue a film badge only to those workers 

who might exceed 10% of the RPGs in effect at that time. This practice resulted in large numbers of 

workers not being monitored for external radiation exposure prior to 1961 (see Figure 6.1.1). 

If a worker’s routine duties and work location remained essentially the same during the 1950s and 

early 1960s, it may be feasible to use his recorded annual doses in the early 1960s to estimate his 

missed dose prior to 1961. Methods are being investigated at present based on department numbers, 

job descriptions, and work locations that might be used to estimate annual doses for other workers 

who were not monitored for external radiation exposure prior to 1961 and did not remain in the same 

jobs during the 1950s and early 1960s.Methods to be considered when there is no recorded dose for 

short periods of time for normally monitored workers have been discussed by Watson et al. (1994). 

Estimates of the missed dose can be made using dose results for co-workers or using the nearby 

recorded dose for the specific worker of interest prior to and after a period of missed dose. 

Regardless of how the missing dose is estimated for non-monitored periods of time, these situations 

do require careful consideration. 

The missed dose for dosimeter results less than the MDL is particularly important for earlier years 

when MDLs were higher and dosimeter exchange was more frequent. NIOSH (2002) describes 

several different options to calculate the missed dose in these situations. One option to estimate a 

claimant-favorable maximum potential dose is to multiply the MDL by the number of zero dose results. 

This will provide an estimate of the maximum missed dose to the worker. The following sections 

consider missed photon dose for dosimeter results less than the MDL according to facility or location, 

dosimeter type, year, and energy range. 

6.5.1.1 Facility or Location 

Information has not been found that is adequate to describe the potential missing photon dose by 

facility or location within the Y-12 Plant. This is particularly true during the early years when the 

missed photon dose was most significant due to the frequent exchange of film dosimeters and the 

higher MDLs. 

6.5.1.2 Dosimeter Type 

The missed photon dose by dosimeter type is discussed in Section 6.3.1 and summarized in Table 

6.3.1-2. The MDLs for the respective Y-12 beta/photon dosimeters are based on results of laboratory 

irradiations, and the actual MDLs for the workplace may be somewhat greater than these values 

because of additional uncertainty in actual field use. Nevertheless, the values provided in Table 

6.3.1-2 are expected to provide reasonable estimates of the missed dose for Y-12 workers. 

6.5.1.3 Year 

Analysis of the missed photon dose by year and dosimeter exchange frequency is discussed in 

Section 6.3.1 and summarized in Table 6.3.1-2. The missed photon dose for workers who were 

unmonitored for external radiation exposures prior to 1961 is also discussed in Section 6.5.1. 

6.5.1.4 Energy Range 

An estimate of the missed dose by energy range may be possible based on the type of facility and 

predominant radiation sources or radionuclides at the facility. The recorded dose from the dosimeter 

response does not typically provide information to estimate discrete energy ranges. It is possible to 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 32 of 47 

examine the energy response characteristics of the multi-element film and TLD dosimeters, but this 

analysis does not recognize the substantial uncertainties present in the workplace associated with 

differing exposure geometries and mixed radiation fields. 

6.5.2 Neutron Missed Dose

There may be significant missed neutron dose at Y-12 because of the very low potential for neutron 

exposure as illustrated by the data in Table 6.3.4.3-2. The neutron missed dose is divided into three 

time periods in the following discussion. The first time period is before 1980 when only NTA film 

dosimeters were used. The second time period is from 1980 to 1989 when the switch from NTA film 

dosimeters to TLNDs was being completed. The third time period is after 1989 when only TLNDs 

were used. The estimated MDLs for these neutron dosimeters are summarized in Table 6.3.1-2. It is 

possible to estimate the missed neutron dose using the MDLs because the neutron dosimeters were 

calibrated with neutron sources that had energies similar to those encountered in the workplace and 

more than 90% of the neutrons to which workers were normally exposed had energies greater than 

the 500-keV threshold of the NTA film dosimeters. There was, of course, no threshold energy for the 

TLNDs as illustrated in Figure 6.3.2.2-1. 

6.5.2.1 Prior to 1980 

The use of NTA films for neutron dosimetry prior to 1980 is well documented in various Y-12 reports. 

As noted above, it is possible to estimate the missed dose using the MDLs. It was also noted 

previously that there are a lot of recorded zeros in the neutron dose data for Y-12 workers for two 

reasons: (1) a worker’s NTA film was not developed and read, or (2) a worker’s NTA film indicated a 

neutron dose equivalent that was less than the film’s MDL, approximately 50 mrem. If the MDL for 

NTA film is used in estimating the missed neutron dose, it should be multiplied by 1.10 for workers in 

Calibration Laboratory and by 1.05 for workers in the Enriched Uranium Storage Area of Building 

9212 and the Nondestructive Analysis Laboratory. It is also possible to estimate the missed neutron 

dose in some facilities by use of neutron-to-photon dose ratios (NIOSH 2002). However, the only 

Y-12 facility where this dose ratio is expected to provide a reasonably reliable estimate of the missed 

neutron dose is the Enriched Uranium Storage Area of Building 9212. For this area, the dose ratio 

was determined by recent PNL measurements to be approximately 1:1. The dose ratios for other 

neutron exposure areas at Y-12 determined from recent PNL measurements were quite large and the 

use of these data could lead to gross overestimates of missed neutron dose if a worker were also 

exposed to pure photon sources at the Calibration Laboratory or the Nondestructive Analysis 

Laboratory. 

6.5.2.2 From 1980 to 1989 

There is a serious gap in the neutron dosimetry information from 1980 to 1989. Y-12 became 

increasingly dependent over the years on ORNL to process the NTA films and to determine their 

neutron doses because of the small numbers of neutron exposed workers. Thus, the neutron 

dosimetry at Y-12 is assumed to be the same as that at ORNL for this time period. All workers at both 

ORNL and Y-12 were provided with a two-element TLD dosimeter for beta-particle and photon 

dosimetry (McLendon 1980a) and workers exposed to neutrons were provided with a separate 

neutron dosimeter (Gupton 1978). This dosimeter was a modification of the film badge dosimeter 

previously used at both ORNL and Y-12. During the switch from film to TLD, the film badge dosimeter 

was modified to hold four TLD chips in a polyethylene mount, 1 mm thick. For workers exposed to 

neutrons, the modified badge contained a combination of two TLD chips and two TLND chips for low 

and intermediate energy neutron dosimetry plus an NTA film for fast neutron dosimetry. The MDL of 

this neutron dosimetry is assumed to about the same as that of the NTA film alone because most of 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 33 of 47 

the neutron dose at Y-12 comes from neutrons above the 500-keV threshold of the NTA film. In the 

mid-1980s, the NTA film was removed because of poor film quality, its large MDL, and the labor 

intensive processing requirements (Berger and Lane 1985). The dosimeter was further modified to 

serve as a neutron albedo dosimeter for neutrons of all energies. This neutron albedo dosimeter 

functioned reasonably well in the workplace neutron fields because it was calibrated using the fast 

fission neutrons from the Health Physics Research Reactor (HPRR) at ORNL (Berger and Lane 

1985). The mean energies of the fast fission neutrons from the HPRR can be varied from 0.56 MeV 

using a concrete shield (or spectrum shifter) to 1.28 MeV with no shielding (or spectrum shifter) 

between the reactor and the dosimeter. Based on very limited data from test studies at the HPRR, it 

is estimated that the MDL for this neutron albedo dosimeter was approximately 20 mrem. 

6.5.2.3 Post 1989 

Since 1989, the neutron dose has been measured using a newly developed albedo-type TLND worn 

on the belt to keep in close contact with the body. The characteristics of this dosimeter are well 

documented (BWXT Y-12 2001) and the MDL to be used in estimating missed dose is 10 mrem (see 

Table 6.3.1-2). 

6.6 UNCERTAINTY IN PHOTON AND NEUTRON DOSE 

For film badges, the MDLs that are quoted in the literature range from about 30 to 50 mrem for 

beta/photon irradiation (Morgan 1961; Parrish 1979; West 1993; Souleyrette 2003) and from 50 to 100 

mrem for neutrons (Morgan 1961; Parrish 1979; Wilson et al. 1990). These are not the expected 

uncertainties at larger photon and neutron dose readings. For example, it was possible to read a 

photon dose of 100 mrem to with ± 15 mrem if the exposure involved photons with energies between 

several hundred keV and several MeV (Morgan 1961). If the exposure involved photons with 

energies less than several hundred keV, the uncertainty was at least twice that for the more energetic 

photons. Thus, the standard error in the recorded film badge doses from photons of any energy is 

estimated here to be ± 30%. The standard error for the recorded dose from beta irradiation was 

essentially the same as that for photon irradiation, but when an unknown mixture of beta and photon 

irradiation was involved the standard error for the dose from beta irradiation was somewhat larger 

than 30% (Morgan 1961). The situation for neutrons was not as favorable as that for photons. With 

NTA films, the estimated standard error was much larger and varied significantly with the energy of 

the neutrons. Thus, the standard error for a neutron dose reading of approximately 100 mrem is 

estimated here to be ± 50%. For the TLD dosimeters used at Y-12 after 1980 and the TLND 

dosimeters used at Y-12 after 1985, the standard errors for a recorded dose reading of 100 mrem or 

more are estimated here to be approximately ± 15% for photons, beta particles, and neutrons. The 

standard errors for TLD and TLND dose measurements less than 100 mrem and for TLD and TLND 

dose measurements in mixed radiation fields would be expected to be somewhat larger. 

6.7 ORGAN DOSE 

Once the photon and neutron doses and their associated standard errors have been calculated for 

each year, the values are then used to calculate organ doses of interest using the NIOSH External 

Dose Reconstruction Implementation Guideline (NIOSH 2002). There are many complexities and 

uncertainties when applying organ dose conversion factors to the adjusted doses of record. Many of 

the factors that affect the recorded dose have already been discussed in the various tables throughout 

this section. Some factors such as backscattering (phantom calibration) and the over response of low 

energy photons would indicate the recorded dose was too high. Other factors such as calibration 

methodology, angular response, low energy threshold, and film fading would result in a recorded dose 

that was too low. As a result differences in film badge design (filtration) and calibration can have both 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 34 of 47 

positive and negative effects on the overall dose comparison to Hp(10). The International 

Commission on Radiation Units and Measurements (ICRU 1988) has indicated that film badge 

dosimeters, while not tissue equivalent, can be used for personnel dosimetry. They also indicate that 

it is more difficult to ensure that the variation in response with energy and angle of incidence with low 

energy. Given the voluminous uncertainties above especially with film badge dosimetry in the 1950-

1980s, a claimant favorable approach is used to estimate organ dose. Since exposure to organ dose 

conversion factors result in a higher organ dose and higher probability of causation, given the 

Radiation Effective Factors of the intermediate energy photons, these dose conversion factors (DCFs) 

will be used to convert recorded film badge doses to organ dose. In the conversion of all recorded 

photon and neutron doses to organ doses, the exposure geometry must also be given careful 

consideration. Some of the more common exposure geometries encountered in the workplace are 

defined as follows: (1) an anterior-posterior (AP or front-to-back) exposure is typical for an individual 

who works in a directional radiation field and faces the source of the radiation source while working, 

(2) rotational exposure is typical of an individual who is constantly turning in a directional radiation 

field while working, and (3) an isotropic exposure is typical of an individual who is working in a highly 

non-directional or omni-directional radiation field. The proposed default options based on claimantfavorable exposure geometries for long-term workers are listed in Table 6.6-1. In Attachment F, the 

use of the parameters presented in Section 6 will be discussed to aid the dose reconstructor in 

preparing dose reconstructions for long-term Y-12 workers. 

Table 6.6-1. Default exposure geometries for calculating organ doses. 

Claim status Job category Exposure geometry Percentage

Likely non-compensable All A-P 100% 

Compensable worker All A-P 

Rotational 

50% 

50% 

Compensable supervisor All A-P 

Isotropic 

50% 

50% 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 35 of 47 

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Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 40 of 47 

GLOSSARY 

IN PREPARATION
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 41 of 47 

ATTACHMENT F 

OCCUPATIONAL EXTERNAL DOSE FOR MONITORED WORKERS 

F.1 OCCUPATIONAL EXTERNAL DOSE 

The information needed to evaluate claims is directed to the technical parameters of the annual 

estimates of the primary organ dose that is calculated from the dosimeter interpreted personal dose 

equivalent, Hp(10), and Hp(0.07) in the case of skin, testicular, and breast cancer. These are used as 

a consistent basis of comparison for all years of Y-12 occupational external dose starting in 1950. 

The primary IREP screen used to input dose parameters is illustrated in Table F-1. The input to these 

fields is obtained from the Y-12 dose of record. The claim provides the primary organ of interest and 

other worker information needed to run the IREP computer program. Guidance to the dose 

reconstruction analysis in selecting the technical external dose parameters to complete the respective 

fields in Table F-1 is provided in the following sections. 

Table F-1. IREP dose parameter input screen. 

Exposure Distribution parameters

# Year Rate Radiation type Type 1 2 3

1 1960 Acute Photon, 30-250 keV Normal 2 2 0 

2 1961 

3 1962 

F.2 YEARS OF EXPOSURE 

The years of exposure should be identified from the Y-12 radiation dose reports. Missed dose is 

calculated for each of the following reasons: (1) the worker was not monitored for external radiation 

exposure before 1961, (2) there is no recorded dose for short periods of time after 1961, and (3) the 

recorded dose during a monitoring period was zero because the dosimeter response was less than 

the MDL. The missed dose should be calculated for a claimant for each year of record as an 

employee unless there are valid reasons for years in which there are no records. 

F.3 EXPOSURE RATE 

Acute is selected for all types of external beta and photon dose and chronic is selected for neutron 

dose (NIOSH 2002). 

F.4 RADIATION TYPE 

F.4.1. Beta and Photon Radiation 

Claimant-favorable assumptions should be made using guidance in Table F-2 for beta particles and 

for photons (x-rays and gamma rays) to assure that dose is not underestimated. The values 

presented in this table are intended to provide a reasonable estimate of parameters used to calculate 

the organ dose without significant numerical error for long-term Y-12 workers in the respective 

facilities. There is no direct evidence to select the specific values shown other than considerations of 

the radiation sources and usual work tasks. In those cases where there is some doubt in the values, 

a range of realistic values should be selected for comparison and the most claimant-favorable option 

selected. 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 42 of 47 

Table F-2. Selection of beta and photon radiation energies and percentages for Y-12 site processes. 

Operations

Y-12 Site Processes Building Begin End

Radiation 

type

Energy 

selection Percent

9203 1947 1951

9206a 1947 1959

9211 1947 1959

Enriched Uranium Product Recovery 

and Salvage Operations

9201-1 1952 1963

Beta 

Photon

> 15 keV 

30-250 keV

100% 

100%

9202 1947 1995

9206a Uranium Chemical Operations and 1947 1995

Weapon Production Operations 9212b 1949 Ongoing

Beta 

Photon

> 15 keV 

30-250 keV

100% 

100%

Special Nuclear Material Receiving 

and Storage

9720-5 1949 Ongoing Photon 30-250 keV 100% 

9201-5 1949 Ongoing

9204-4 1949 Ongoing

Uranium Forming and Machining for 

Weapon Component Operations

9215 1950 Ongoing

Beta 

Photon

> 15 keV 

30-250 keV 

100% 

100% 

9201-5 1949 Ongoing

9204-4 1949 Ongoing

9766 1949 ? Depleted Uranium Process Operations

9998 1949 Ongoing

Beta 

Photon > 15 keV 

30-250 keV 

> 250 keV

100% 

50% 

50%

Final Weapon Component Assembly 9204-2 1952 Ongoing

Operations 9204-2E 1952 Ongoing

Beta 

Photon

> 15 keV 

30-250 keV

100% 

100%

ORNL 86-Inch Cyclotron 9201-2 1950 ? Photon 30-250 keV 

>250 keV 

50% 

50% 

Chemical Assay and Mass 

Spectrometry Laboratories

9203 1947 Ongoing Photon

Radiographic Laboratory 9201-1 1947 Ongoing Photon

Calibration Laboratory 9983 1949 Ongoing Photon

Weapon Component Assay Laboratory 9995 1952 Ongoing Photon

Nondestructive Assay Laboratory 9720-5 1980 Ongoing Photon 

Specific to 

radiation source 

Photon default 

values: 

20-250 keV 

>250 keV

50% 

50%

West End Waste Treatment Facility

9616-7 1984 Ongoing Beta 

Photon

> 15 keV 

30-250 keV 

>250 keV

100% 

50% 

50%

a. Building 9206 Complex includes Buildings 9768, 9720-17, 9409-17, 9510-2, 9767-2, and the east and west tank farm pits. 

b. Building 9212 Complex includes Buildings 9809, 9812, 9818, 9815, and 9980. 

F.4.2. Neutron Radiation 

The default neutron dose distributions by energy for each of the neutron exposure areas at Y-12 are 

summarized in Table F-3. 

Table F-3. Selection of neutron energies and percentages for Y-12 site facilities. 

Operations

Y-12 site facility Building Begin End

Neutron 

energy

Default dose 

fraction (%)

Calibration Laboratory 9983 1949 Ongoing 0.1-2 MeV 

2-14 MeV 

57% 

43% 

Enriched Uranium Storage Area 9212 1949 Ongoing 0.1-2MeV 

2-14 MeV 

79% 

21% 

Nondestructive Assay Laboratory 9720-5 1980 Ongoing 0.1-2 MeV 

2-14 MeV 

23% 

77% 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 43 of 47 

F.5 ADJUSTMENTS TO RECORDED DOSE 

F.5.1. Parameter #1 

Selection of the distribution parameters in Table F-1 involves the adjustments to the dose of record for 

missed dose prior to entry into the IREP input screen. The selection of a normal distribution for the 

“Type” determines the definition of Parameters #1 and #2. For a normal distribution, Parameter #3 is 

not used and Parameter #1 is the mean of the distribution of recorded dose for each year of 

monitoring. 

F.5.2. Adjustment to Recorded Photon Dose 

Adjustments to the Y-12 reported photon dose are necessary to arrive at a claimant-favorable dose 

considering uncertainties associated primarily with the complex workplace radiation fields and 

exposure geometries. Henderson (1991) identified such problems for workers performing waist-level 

handling jobs in the DU process areas of Y-12. Examples of waist-level handling jobs are unloading 

and sorting of DU scrap materials, shearing of larger pieces of scrap materials, cleaning of the scrap 

materials, crucible loading during the melting and casting operations, and materials sampling. It is 

now a practice to instruct workers performing these operations to wear their beta/photon dosimeters 

at the waist, but many of these workers may have worn their dosimeters on the collar in the past. The 

photon dose correction summarized in Table F-4 is necessary to calculate an adjusted photon dose 

that is claimant favorable prior to 1991. From 1991 to present, no correction in needed because the 

recorded dose is Hp(10) equivalent. To determine when to make such adjustments, the dose 

reconstructor must depend on information about routine duties and work locations that are contained 

in the Computer Assisted Telephone Interview (CATI) file for a claimant. 

Table F-4. Adjustments to reported Y-12 deep photon dose. 

Parameter Description

Period Prior to 1/1/1991 

Dosimeters All beta/photon dosimeters 

Facilities Depleted uranium process operations 

Workers Waist-level metal handling operators 

Adjustment to 

recorded dose 

Multiply reported deep photon dose by a factor of 1.34 

to estimate Hp(10) 

F.5.3. Adjustments to Recorded Neutron Dose 

The Y-12 facility incorporated the energy variation of the dose equivalent from neutrons into their 

calibration methodology. Thus, the recorded dose equivalent is a combination of all neutron energies. 

In order to calculate the neutron dose input to IREP (see Table F-1), the recorded neutron dose must 

be separated into neutron energy groups as shown in Table F-3 and subsequently converted into 

ICRP 60 methodology (ICRP 1990). The dose fractions by neutron energy group and the associated 

ICRP 60 correction factors for the various neutron exposure areas at Y-12 are summarized in Table 

F-5. As an example, consider security personnel who inventory fissile material in the Enriched 

Uranium Storage Area of Building 9212 and assume that such a person receives a neutron dose of 

100 mrem. The corrected neutron dose is 151 mrem for neutrons with energies between 0.1-2 MeV 

and 28 mrem for neutrons with energies between 2-14 MeV. Thus, the total corrected neutron dose is 

a total of 179 mrem. These corrections should be applied to both recorded dose and missed dose. 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 44 of 47 

F.5.4. Unmonitored Photon and Neutron Dose 

Missed photon and neutron would occur where there is no recorded dose because the worker was not 

monitored or the dose is unavailable for a short period of time because a film was either lost or 

damaged while being processed. Before 1961, the policy at Y-12 was to issue a film badge only to 

those workers who might exceed 10% of the RPGs in effect at that time. This practice resulted in 

large numbers of workers not being monitored for external radiation exposure prior to 1961 (see 

Figure 6.1.1). 

Table F-5. Neutron dose fractions and associated ICRP 60 correction factors for Y-12 site facilities. 

Operations

Y-12 facilities Building Begin End Neutron energy

Neutron dose 

fraction

ICRP 60 correction 

factors

Calibration Laboratory 9983 1949 Ongoing 0.1-2 MeV 

2-14 MeV

0.57 

0.43

1.09 

0.57

Enriched Uranium 

Storage Area

9212 1949 Ongoing 0.1-2MeV 

2-14 MeV

0.79 

0.21

1.51 

0.28

Nondestructive 

Analysis Laboratory

9720-5 1980 Ongoing 0.1-2 MeV 

2-14 MeV

0.23 

0.77 

0.44 

1.02 

If a worker’s routine duties and work location remained essentially the same during the 1950s and 

early 1960s, it may be feasible to use his recorded annual doses in the early 1960s to estimates his 

missing dose prior to 1961. Methods are being investigated at present using department numbers, 

job descriptions, and work locations that might be used to estimate annual doses for other workers 

who were not monitored for external radiation exposure prior to 1961 and did not remain in the same 

jobs during the 1950s and early 1960s. 

Methods to be considered when there is no recorded dose for short periods of time for normally 

monitored workers have been discussed by Watson et al. (1994). Estimates of the missed dose can 

be made using dose results for coworkers or using nearby recorded dose for the specific worker of 

interest prior to and after a period of missed dose due to a film that was either lost or damaged while 

being processed. Regardless of how the missing dose is estimated for non-monitored periods of time, 

these situations do require careful consideration. 

F.5.5. Missing Photon Dose 

Missing photon dose also occurs when the recorded dose is zero because the dosimeter response 

was less than the MDL. This kind of missed dose is most important for earlier years when MDLs were 

higher and dosimeter exchange was more frequent. NIOSH (2002) guidance should be followed to 

calculate the missing photon dose by using a claimant-favorable maximum potential missed dose. 

This is calculated by multiplying the MDL by the number of zero dose results to estimate the 

maximum potential missed dose. The following sections discuss the missed photon dose corrections 

according to facility or location, dosimeter type, year, and energy range. 

F.5.5.1. Facility or Location 

Information has not been found that is adequate to describe the potential missing photon dose by 

facility or location within Y-12. This is particularly true during the early years when the missed photon 

dose was most significant due to the frequent exchange of film dosimeters and their higher MDLs. 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 45 of 47 

F.5.5.2. Dosimeter Type 

The missed photon dose by dosimeter type is summarized in Table F-6. The MDLs for the respective 

Y-12 beta/photon dosimeters are based on results of laboratory irradiations. The actual MDLs for the 

workplace may be somewhat greater than these values because of additional uncertainty in actual 

field use. Nevertheless, the values provided in Table F-6 are expected to provide reasonable 

estimates of the missed dose for Y-12 workers. 

F.5.5.3 Year 

Analysis of the missed photon dose by year and dosimeter exchange frequency is summarized in 

Table F-6. The missed photon dose for workers who were unmonitored for external occupational 

radiation exposure prior to 1961 is also discussed in Section F.5.4. 

F.5.5.4. Energy Range 

An estimate of the missed dose by energy range may be possible based on the type of facility and 

predominant radiation sources or radionuclides at the facility. The recorded dose from the dosimeter 

response does not typically provide information to estimate discrete energy ranges. It is possible to 

examine the energy response characteristics of the multi-element film and TLD dosimeters, but this 

analysis does not recognize the substantial uncertainties present in the workplace associated with 

differing exposure geometries and mixed radiation fields. 

Table F-6. Dosimeter type, period of use, exchange frequency, laboratory minimum detectable limit, 

and maximum annual missed dose. 

Dosimeter Period

Exchange 

frequency

Laboratory 

MDL (mrem)

Maximum annual 

missed dose (mrem)

Beta/photon dosimeters 

Pocket ionization chamber 1948-1950 Daily < 5 1,300 

Weekly < 5 260 

Two-element film badge 1948-1958 Weekly 40 2,080 

1958-1961 Monthly 40 480 

Four-element film badge 1961-1980 Quarterly 40 160 

Two-element TLD dosimeter 1980-1989 Quarterly 20 80 

Four-element TLD dosimeter 1989-Present Quarterly 10 40 

Neutron dosimeters 

Biweekly < 50 1,300 

Monthly < 50 600 

NTA film 1948-1980 

Quarterly < 50 200 

Combination NTA film and TLND 1980-1985 Quarterly < 50 200 

dosimeter 1985-1989 Quarterly 20 80 

TLND dosimeter 1989-Present Quarterly 10 40 

F.5.6. Neutron Missed Dose 

The estimated MDLs for the neutron dosimeters used at Y-12 are summarized in Table F-6. It is 

possible to calculate the missed neutron dose at Y-12 using the MDLs because the neutron 

dosimeters were calibrated with neutron sources that had energies similar to those encountered in the 

workplace and more than 90% of the neutrons to which workers were normally exposed had energies 

greater than the 500-keV threshold of the NTA film dosimeters. If the MDL for NTA film is used in 

estimating the missed neutron dose, it should be multiplied by 1.10 for workers in the Calibration 

Laboratory and by 1.05 for workers in the Enriched Uranium Storage Area of Building 9212 and the 

Nondestructive Analysis Laboratory. It is also possible to estimate missed neutron dose in some 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 46 of 47 

facilities by use of neutron-to-photon dose ratios (NIOSH 2002). However, the only Y-12 facility 

where a neutron-to-photon dose ratio is expected to provide a reasonably reliable estimate of the 

missed dose is for workers in the Enriched Uranium Storage Area of Building 9212. For this area, the 

neutron-to-dose ratio was approximately 1:1. The neutron-to-dose ratios for other neutron exposure 

areas at Y-12 were quite large and the use of these data could lead to gross overestimates of missed 

neutron dose if an individual was also exposed to pure photon sources during the course of their work 

at either the Calibration Laboratory or the Nondestructive Analysis Laboratory. 

F.5.7 Organ Dose Equivalent 

Once the adjusted photon and neutron doses have been calculated for each year, the values are used 

to calculate organ doses of interest using the NIOSH External Dose Reconstruction Implementation 

Guideline (NIOSH 2002). There are many complexities and uncertainties when applying organ dose 

conversion factors to the adjusted doses of record. Many of the factors that affect the recorded dose 

have already been discussed in the various tables throughout this section. Some factors such as 

backscattering (phantom calibration) and the over response of low energy photons would indicate the 

recorded dose was too high. Other factors such as calibration methodology, angular response, low 

energy threshold, and film fading would result in a recorded dose that was too low. As a result 

differences in film badge design (filtration) and calibration can have both positive and negative effects 

on the overall dose comparison to Hp(10). The International Commission on Radiation Units and 

Measurements (ICRU 1988) has indicated that film badge dosimeters, while not tissue equivalent, can 

be used for personnel dosimetry. They also indicate that it is more difficult to ensure that the variation 

in response with energy and angle of incidence with low energy. Given the voluminous uncertainties 

above especially with film badge dosimetry in the 1950-1980s, a claimant favorable approach is used 

to estimate organ dose. Since exposure to organ dose conversion factors result in a higher organ 

dose and higher probability of causation, given the Radiation Effective Factors of the intermediate 

energy photons, these DCFs will be used to convert recorded film badge doses to organ dose. In the 

conversion of all recorded photon and neutron doses to organ doses, the exposure geometry must 

also be given careful consideration. The proposed default options based on claimant-favorable 

exposure geometries for long-term workers are listed in Table F-7. 

F.5.8 Parameter #2 

Parameter #2 is the standard deviation of the normal distribution for the organ dose. The individual 

dose result for each dosimeter exchange period will be available to calculate the mean and standard 

deviation for each year. If not available, the adjusted organ dose can be used for each year and a 

default standard deviation value used for parameter #2. 

Table F-7. Default exposure geometries for calculating organ dose. 

Claim status 

Job 

category 

Exposure 

geometry Percentagea

Non-compensable All A-P 100% 

Compensable worker All A-P 

Rotational 

50% 

50% 

Compensable supervisor All A-P 

Isotropic 

50% 

50% 

a. Apply this percentage to the dose conversion factor in Appendix B of NIOSH 

(2002) to arrive at the total organ dose equivalent from the adjusted recorded 

dose. 
Effective Date: 11/19/2003 Revision No. 00 Document No. ORAUT-TKBS-0014-6 Page 47 of 47 

F.5.9 Organ Dose Conversion Factors 

A detailed discussion of the conversion of measured dose to organ dose equivalent is provided in 

Appendix A of NIOSH (2002). Appendix B of NIOSH (2002) contains the appropriate dose conversion 

factors (DCFs) for each organ, radiation type, and energy range based on the type of monitoring 

performed. In some cases, simplifying assumptions are appropriate.